{{Adams | number = ML15176A649 | issue date = 06/19/2015 | title = Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2 | author name = Shea J W | author affiliation = Tennessee Valley Authority | addressee name = | addressee affiliation = NRC/Document Control Desk, NRC/NRR | docket = 05000327, 05000328 | license number = DPR-077, DPR-079 | contact person = | case reference number = CNL-15-128, SQN-TS-11-10 | package number = ML15176A678 | document type = Letter, Technical Specifications | page count = 2214 }}

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{{#Wiki_filter:Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37 402 CNL-15-128 June 19, 2015 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328 10 CFR 50.90

Subject:

Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10)-Supplement 2

References:

1. TVA Letter to NRC, "Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-1 0)," dated November 22, 2013. (ADAMS Accession No. ML 13329A717) 2. TVA Letter to NRC, "Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-1 Supplement 1," dated December 16,2014. (ADAMS Accession No. ML 14350B364) By letter dated November 22, 2013, Tennessee Valley Authority (TVA) requested a license amendment to revise the current Technical Specifications for Sequoyah Nuclear Plant (SQN), Units 1 and 2, to the Improved Technical Specifications (ITS) consistent with the Improved Standard Technical Specifications described in NUREG-1431, "Standard Technical Specifications-Westinghouse Plants," Revision 4.0 (Reference 1 ). By letter dated December 16, 2014 (Reference 2), TVA provided a supplement to the ITS license amendment request (LAR). The supplement provided information regarding the revised SQN fuel handling accident radiological consequences analysis using the alternative source term. The purpose of this letter is to supplement the original ITS LAR (Reference 1 ). Specifically, this letter complements and revises the original LAR based on responses to NRC staff requests for additional information (RAis). The RAis and the associated responses are posted on a publicly available website, the NRC and SQN ITS Conversion Website (http://www.excelservices.com), and are hereby docketed by submittal of this letter.

U.S. Nuclear Regulatory Commission CNL-15-128 Page 2 of 4 June 19, 2015 This supplement contains: 1) proposed changes to the original ITS LAR resulting from TVA responses to the ITS LAR RAis; 2) proposed changes resulting from TV A self-identified issues discovered during review of the ITS LAR; and 3) docketed submittal of NRC staff RAis, TVA responses and NRC staff RAI closures posted as of May 31, 2015. Enclosure 1, "Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-1 0)-Revision 1 ," provides a revision of Enclosures 2 (Volumes 1, 3, 4, 5, 6, 7, 9, 10, 15 and 16), 5, 6, 8, and 9 of the original ITS LAR with changes annotated on the affected pages. ITS LAR revisions associated with a TVA response to an RAI are identified on the revised page with a red text box indicating the applicable RAI number (e.g., KAB044). ITS LAR revisions associated with a TVA self identified issue are identified on the revised page with a red text box with an Sll indicator. Volumes 1, 3, 4, 5, 6, 7, 9, 10, 15 and 16 of Enclosure 2 of the original ITS LAR have been included because there are no open RAis associated with them. Additionally, Enclosure 1 contains a revision to Enclosure 8 of the original ITS LAR. Revised Enclosure 8, "Regulatory Commitments," contains two new commitments associated with responses to RAis MEH-006 and RPG-014. There are thirteen commitments associated with the ITS LAR. The commitments associated with the ITS LAR are only those contained in the revised Enclosure 8. Enclosure 2, "SQN Self-Identified Issues," provides a list of self-identified issues discovered during review of the ITS LAR (Reference 1 ). The list provides a brief description of each self-identified issue, the ITS Section affected by the issue, and the page numbers for affected pages. Enclosure 3, "SQN ITS Conversion RAI Database," contains the NRC staff RAis and the associated TVA responses. Each RAI/response includes the question asked by the NRC staff reviewer, the TVA response, proposed changes to pages contained in the ITS LAR (Reference 1 }, any attached supporting documentation, and RAI closure documentation as of May 31, 2015. The information provided by this supplement to the original ITS LAR does not change the intent or the justification for the requested ITS license amendment. TVA has further determined that this supplement does not affect the basis for concluding that the proposed license amendment does not involve a Significant Hazards Consideration. As such, the 10 CFR 50.92 evaluation provided in the November 22, 2013, ITS LAR remains valid. In addition, the ITS LAR, including this supplement, continues to be exempt from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). The SQN Plant Operations Review Committee has reviewed this supplemental information and determined that operation of SQN in accordance with the Technical Specifications as proposed in the original ITS LAR and this supplement, will not endanger the health and safety of the public. U.S. Nuclear Regulatory Commission CNL-15-128 Page 3 of 4 June 19, 2015 Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the enclosure to the Tennessee State Department of Environment and Conservation. If there are any questions or if additional information is needed, please contact Mr. Tom Hess at 423-751-3487. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 19th day of June 2015. fully,

Enclosures:

Enclosure 1 -Sequoyah Nuclear Plants, Units 1 and 2 Technical Enclosure cc (Enclosure): Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-1 0) -Revision 1 Enclosure 2 -SQN Self-Identified Issues Enclosure 3 -SON ITS Conversion RAI Database NRC Regional Administrator-Region II NRC Senior Resident Inspector-Sequoyah Nuclear Plant Director, Division of Radiological Health-Tennessee State Department of Environment and Conservation NRC Project Manager-Sequoyah Nuclear Plant ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 - ENCLOSURE 2 VOLUME 1 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Revision 0

APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS i CONTENTS Page 1. INTRODUCTION ......................................................................................................... 1 2. SELECTION CRITERIA ............................................................................................... 2 3. PRA INSIGHTS ........................................................................................................... 5

4. RESULTS OF APPLICATION OF SELECTION CRITERIA ........................................ 8

5. REFERENCES ........................................................................................................... 9 ATTACHMENT

1. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2

APPENDIX A. JUSTIFICATION FOR SPECIFICATION RELOCATION APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 1 of 8 1. INTRODUCTION The purpose of this document is to confirm the results of the Westinghouse Owners Group application of the Technical Specification selection criteria on a plant specific basis for the Sequoyah Nuclear Plant (SQN) Unit 1 and Unit 2. The Tennessee Valley Authority (hereinafter TVA) has reviewed the application and confirmed the applicability of the selection criteria to each of the Technical Specifications utilized in report WCAP-11618, "Methodically Engineered Restructured and Improved Technical Specifications, MERITS Program - Phase II Task 5, Criteria Application" (Reference 1) including Addendum 1, NRC Staff Review of NSSS Vendor Owners Groups Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications, Wilgus/Murley letter dated May 9, 1988 and as revised in NUREG-1431, Revision 4.0 "Standard Technical Specifications, Westinghouse Plants" (Reference 2) and applied the criteria to each of the current SQN Technical Specifications. Additionally, in accordance with the NRC Final Policy Statement (Reference 3), this confirmation of the application of selection criteria includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in Reference 1, as applicable to SQN. APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 2 2. SELECTION CRITERIA TVA has utilized the selection criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 22, 1993 (Reference 3) to develop the results contained in the attached matrix. PRA insights as used in the Westinghouse Owners Group submittal were utilized, confirmed by TVA, and are discussed in the next section of this report. The selection criteria and discussion provided in Reference 3 are as follows: Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary: Discussion of Criterion 1: A basic concept in the adequate protection of the public health and safety is the prevention of accidents. Instrumentation is installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safely, thus reducing the likelihood of a loss-of-coolant accident. This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage. This criterion should not, however, be interpreted to include instrumentation to detect precursors to reactor coolant pressure boundary leakage or instrumentation to identify the source of actual leakage (e.g., loose parts monitor, seismic instrumentation, valve position indicators). Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier: Discussion of Criterion 2: Another basic concept in the adequate protection of the public health and safety is that the plant shall be operated within the bounds of the initial conditions assumed in the existing design basis accident and transient analyses and that the plant will be operated to preclude unanalyzed transients and accidents. These analyses consist of postulated events, analyzed in the Final Safety Analysis Report (FSAR), for which a structure, system, or component must meet specified functional goals. These analyses are contained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as Condition II, III, or IV events APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 3 (ANSI N18.2) (or equivalent) that either assume the failure of or present a challenge to the integrity of a fission product barrier. As used in Criterion 2, process variables are only those parameters for which specific values or ranges of values have been chosen as reference bounds in the design basis accident or transient analyses and which are monitored and controlled during power operation such that process values remain within the analysis bounds. Process variables captured by Criterion 2 are not, however, limited to only those directly monitored and controlled from the control room. These could also include other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g, moderator temperature coefficient and hot channel factors).

The purpose of this criterion is to capture those process variables that have initial values assumed in the design basis accident and transient analyses, and which are monitored and controlled during power operation. As long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure/low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients. Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier: Discussion of Criterion 3: A third concept in the adequate protection of the public health and safety is that in the event that a postulated design basis accident or transient should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequences of the design basis accident or transient. Safety sequence analyses or their equivalent have been performed in recent years and provide a method of presenting the plant response to an accident. These can be used to define the primary success paths. A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's design basis accident and transient analyses, APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 4 as presented in Chapters 6 and 15 of the plant's Final Safety Analysis Report (or equivalent chapters). Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria. It is the intent of this criterion to capture into Technical Specifications only those structures, systems, and components that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path for a particular mode of operation does not include backup and diverse equipment (e.g., rod withdrawal block which is a backup to the average power range monitor high flux trip in the startup mode, safety valves which are backup to low temperature overpressure relief valves during cold shutdown). Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety: Discussion of Criterion 4: It is the Commission's policy that licensees retain in their Technical Specifications LCOs, Action statements and Surveillance Requirements for the following systems (as applicable), which operating experience and PRA have generally shown to be significant to public health and safety and any other structures, systems, or components that meet this criterion:

Reactor Core Isolation Cooling/Isolation Condenser,
Residual Heat Removal,
Standby Liquid Control, and
Recirculation Pump Trip. The Commission recognizes that other structures, systems, or components may meet this criterion. Plant and design-specific PRA's have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report Design Basis Accident or Transient analyses. It is the intent of this criterion that those requirements that PRA or operating APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 5 experience exposes as significant to public health and safety, consistent with the Commission's Safety Goal and Severe Accident Policies, be retained or included in Technical Specifications. The Commission expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant specific PRA or risk survey and any available literature on risk insights and PRAs. This material should be employed to strengthen the technical bases for those requirements that remain in Technical Specifications, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk. Similarly, the NRC staff will also employ risk insights and PRAs in evaluating Technical Specifications related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements. 3. PRA INSIGHTS

Introduction and Objectives Reference 3 includes a statement that NRC expects licensees to utilize any plant specific PRA or risk survey and any available literature on risk insights and PRAs to strengthen the technical bases for these requirements that remain in Technical Specifications and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk. Those Technical Specifications proposed as being relocated to other plant controlled documents will be maintained under programs subject to the 10 CFR 50.59 review process. These Relocated Specifications have been compared to a variety of PRA material with two purposes: 1) to identify if a Specification component or topic is addressed by PRA; and 2) if addressed, to judge if the Relocated Specification component or topic is risk-important. The intent of the PRA review was to provide an additional screen to the deterministic criteria. This review was accomplished in the generic Westinghouse Owners Group submittal WCAP-11618 and Addendum 1 to WCAP-11618 (Reference 1). The results of this generic review have been confirmed by TVA for the applicable SQN Specifications to be relocated. APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 6 Assumptions and Approach The WCAP-11618 evaluation of the risk impact of the Technical Specifications that are relocation candidates was based on the following: a. It was assumed that any of the Technical Specifications that were to be relocated would be transferred to other documents subject to control by the utility under the 10 CFR 50.59 process. b. The risk criteria used in determining the disposition of a Technical Specification were the following: 1. If the Technical Specification contained constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk, it should be retained; 2. If the Technical Specification included items involved in one of these dominant sequences but had an insignificant impact on the probability or severity of that sequence, it was proposed to be relocated to another controlled document; and 3. If the Technical Specification was not involved in risk dominant sequences, it was proposed to be relocated to another controlled document. c. The measures related to risk used in this evaluation were core damage frequency and off-site health effects. These measures were consistent with the Final Policy Statement on Technical Specifications and the Safety Goal and Severe Accident Policy Statements. d. The criteria used to determine if a sequence was risk dominant was the following: For core damage, any sequence whose frequency was commonly found to be greater than 1 X 10-6 per reactor year was maintained as a possible dominant sequence as a conservative first cut. This was roughly 2% of the total core damage frequency of 5 X 10-5 for typical PRAs. Each specific sequence identified in the screening of the Technical APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 7 Specifications was evaluated against the above conservative criterion to determine if it was risk dominant. For off-site health effects, any sequence whose frequency of serious radioactive release was commonly found to be greater than 1 x 10-7 per reactor year was considered to be a dominant risk sequence for the purposes of WCAP-11618. This criterion was in agreement with the NRC position in the Safety Goal Policy for a goal of 1 X 10-6 for a total frequency of severe off-site release, and no greater than 1 X 10-7 for an individual sequence. e. Included in Section 4.0 of WCAP-11618, were two tables (Tables 3 and 4) which contained representative sequences for all identified types of initiating events considered in formal risk assessments for two types of reference plants. Table 3 was representative of a plant with a large dry containment and Table 4 contained the dominant accident sequences for a plant with a subatmospheric containment. These lists were based on industry PRAs and were reviewed for consistency with NRC sponsored PRA programs. The results were found to be consistent. Systems identified in Tables 3 and 4 of Section 4.0 of WCAP-11618 that contributed significantly to risk as defined in Paragraph d above were listed in Tables 3A, 3B, 4A and 4B of Section 4.0. These identified systems as well as sequences and the risk dominant initiating events from Tables 3 and 4 which were involved in typical dominant core damage and serious release sequences from formal risk assessments were used to screen the requirements of the Technical Specifications reviewed. Those Technical Specifications whose requirements were relevant to these systems, sequences, and initiating events were further evaluated for risk dominance. The remaining Technical Specifications were evaluated on the basis of risk insights from references listed in Section 4.0, Appendix B of WCAP-11618. If the requirements of a Technical Specification were not found to be modeled in any reference and no significant issues were identified from a review of the risk insights, the conclusion was that it did not contain constraints of prime importance to limiting the likelihood or severity of sequences that are commonly found to dominate risk. APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 8 4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the SQN Technical Specifications. The following Summary Disposition Matrix is a summary of that application indicating which Specifications are being retained or relocated, the criteria for inclusion, if applicable, the NRC results of the criteria application as expressed in the NRC Staff Review of NSSS Vendor Owners Groups Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications, Wilgus/Murley letter dated May 9, 1988, and any necessary explanatory notes. Discussions that document the rationale for the relocation of each Specification which failed to meet the selection criteria are provided in Appendix A, except as noted in the Summary Disposition Matrix. APPLICATION OF SELECTION CRITERIA TO THE SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS 9 5. REFERENCES 1. WCAP-11618 (and Addendum 1), "Methodically Engineered Restructured and Improved Technical Specifications, MERITS Program-Phase II Task 5, Criteria Application," November 1987. 2. NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revision 4.0, April 2001. 3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132). ATTACHMENT 1 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 1 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 1.0 1.3 1.7 1.30 DEFINITIONS1.1 5.5.1 3.6.3 3.6.1/3.6.3 3.6.7/3.6.10 YES This section provides definitions for several defined terms used throughout the remainder of Technical Specifications. They are provided to improve the meaning of certain terms. As such, direct application of the Technical Specification selection criteria is not appropriate. However, only those definitions for defined terms that remain as a result of application of the selection criteria, will remain as definitions in this section of Technical Specifications. 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.0 2.1 Safety Limits 2.1 2.1.1 Reactor Core-The combination of Thermal Power, Pressurizer Pressure and Highest Operating Loop Coolant Tavg Shall not exceed the limits of the COLR 2.1.1 YES Application of Technical Specification selection criteria is not appropriate. However, Safety Limits will be included in Technical Specifications as required by 10 CFR 50.36. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 2 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 2.1.2 Reactor Coolant System Pressure- Reactor Coolant System Pressure shall not exceed 2735 2.1.2 YES Application of Technical Specification selection criteria is not appropriate. However, Safety Limits will be included in Technical Specifications as required by 10 CFR 50.36. 2.2 Limiting Safety System Settings 2.2.1 Reactor Protection System Setpoints 3.3.1 YES-3 The RTS LSSS have been included as part of the RPS instrumentation Specification, which has been retained since the Functions either actuate to mitigate consequences of design basis accidents and transients or are retained as directed by the NRC as the Functions are part of the RTS. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 3 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS - APPLICABILITY 3.03.0.1 Operational Modes LCO 3.0.1 YES This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1431, Revision 4. 3.0.2 Noncompliance LCO 3.0.2 YES Same as above. 3.0.3 Generic Actions LCO 3.0.3 YES Same as above. 3.0.4 Entry into Operational Modes LCO 3.0.4 YES Same as above. 3.0.5 Operability Exception 3.8.1 YES The application of Technical Specification selection criteria is not appropriate. However, this exception to the definition of OPERABILITY has been included as part of the Required Actions in ITS 3.8.1. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 4 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3.0.6 Actions Exceptions LCO 3.0.5 YES This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1431, Revision 4. 3.0.7 Snubbers LCO 3.0.8 YES 4.0.1 Operational Modes SR 3.0.1 YES This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1431, Revision 4. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 5 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 4.0.2 Time of Performance SR 3.0.2 YES Same as above. 4.0.3 Noncompliance SR 3.0.3 YES Same as above. 4.0.4 Entry into Operational Modes SR 3.0.4 YES Same as above. 4.0.5 ASME Code Class 1, 2, and 3 Components 5.5.5 5.5.6 YES This Specification is actually a Surveillance Requirement which has been retained in the Administrative Controls programs for Inservice Testing. 3/4.1 REACTIVITY CONTROL SYSTEMS 3.1 3/4.1.1 Boration Control 3.1.1.1 SHUTDOWN MARGIN- Tavg Greater Than 200oF 1.0 3.1.1 3.1.2 3.1.4 3.1.6 YES-2 3.1.1.2 SHUTDOWN MARGIN- Tavg Less Than or Equal to 200oF 1.0 3.1.1 3.1.4 YES-2 3.1.1.3 Moderator Temperature Coefficient 3.1.3 YES-2 3.1.1.4 Minimum Temperature for Criticality 3.4.2 YES-2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 6 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.1.2 Boration Systems (deleted) 3/4.1.3 Movable Control Assemblies 3.1.3.1 Group Height 3.1.4 YES-2 3.1.3.2 Position Indicator Systems-Operating 3.1.7 YES-2 3.1.3.4 Rod Drop Time 3.1.4 YES-2 This Specification has been incorporated as a Surveillance Requirement (SR 3.1.4.3) in ITS 3.1.4. 3.1.3.5 Shutdown Rod Insertion Limit 3.1.5 YES-2 3.1.3.6 Control Rod Insertion Limits 3.1.6 YES-2 3/4.2 POWER DISTRIBUTION LIMITS 3.2 3/4.2.1 Axial Power Imbalance 3.2.1 Axial Flux Difference(AFD) 3.2.3 YES-2 3/4.2.2 Heat Flux Hot Channel Factor - FQ(X,Y,Z) 3.2.2 Heat Flux Hot Channel Factor - FQ(X,Y,Z) 3.2.1YES-2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 7 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FH (X,Y) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FH(X,Y) 3.2.2 YES-2 3/4.2.4 Quadrant Power Tilt Ratio 3.2.4 Quadrant Power Tilt Ratio 3.2.4 YES-2 3/4.2.5 DNB Parameters 3.2.5 DNB Parameters 3.4.1 YES-2 3/4.3 INSTRUMENTATION 3.3 3/4.3.1 Reactor Trip System Instrumentation 3.3.1.1 U1 3.3.1 U2 Reactor Trip System Instrumentation 3.3.1 3.3.2 3.3.9 YES-3 3/4.3.2 Engineered Safety Feature Actuation System Instrumentation 3.3.2.1 U1 3.2.1 U2 Engineered Safety Feature Actuation System Instrumentation 3.3.2 3.3.6 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 8 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.3.3 Monitoring Instrumentation 3.3.3.1 Radiation Monitoring Instrumentation 3.3.6 3.3.7 3.3.8 YES-3 Instrument 1 Area Monitors Instrument 1.a Fuel Storage Pool Area Emergency Ventilation System Actuation 3.3.8 YES-3 Instrument 2 Process Monitors Instrument 2.a Containment Purge Air 3.3.6 YES-3 Instrument 2.b.ii Containment Particulate Activity RCS Leakage Detection 3.4.15 YES-1 Instrument 2.c Control Room Isolation 3.3.7 YES-3 3.3.3.5 Remote Shutdown Instrumentation 3.3.4 YES-4 3.3.3.7 Accident Monitoring Instrumentation 3.3.3 5.6.5 YES-3 3.3.3.10 Explosive Gas Monitoring Instrumentation Relocated NO See Appendix A, Page 1. 3.3.3.11 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3.5 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 9 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.4 REACTOR COOLANT SYSTEM 3.4 3/4.4.1 Reactor Coolant Loops and Coolant Circulation 3.4.1.1 Startup and Power Operation 3.4.4 YES-2 3.4.1.2 Hot Standby 3.4.5 YES-3 3.4.1.3 Shutdown 3.4.6 YES-4 3.4.1.4 Cold Shutdown 3.4.7 3.4.8 YES-4 3/4.4.3 Safety and Relief Valve - Operating 3.4.3.1 Safety Valves-Operating 3.4.10 YES-3 3.4.3.2 Relief Valves -Operating 3.4.11 YES-3 3/4.4.4 Pressurizer 3.4.4 Pressurizer 3.4.9 YES-2 3/4.4.5 Steam Generators 3.4.5 Steam Generators 3.4.17 5.5.9 YES-2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 10 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.4.6 Reactor Coolant System Leakage 3.4.6.1 Leakage Detection Instrumentation 3.4.15 YES-1 3.4.6.2 Operational Leakage 3.4.13 YES-2 3.4.6.3 Reactor Coolant System Pressure isolation Valve Leakage 3.4.14 YES-2 3/4.4.8 Specific Activity 3.4.8 Specific Activity 3.4.16 YES-2 3/4.4.9 RCS Pressure and Temperature (PT)Limits 3.4.9.1 RCS Pressure and Temperature (PT)Limits 3.4.3 YES-2 3/4.4.12 Low Temperature Over Pressure Protection (LTOP) System 3.4.12 Low Temperature Over Pressure Protection (LTOP) System 3.4.12 YES-2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 11 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.5 EMERGENCY CORE COOLING SYSTEMS(ECCS) 3.5 3/4.5.1 Accumulators 3.5.1 Cold Leg Injection Accumulators 3.5.1 YES-3 3/4.5.2 ECCS Subsystems - Operating 3.5.2 ECCS Subsystems - Operating 3.5.2 YES-3 3/4.5.3 ECCS Subsystems - Shutdown 3.5.3 ECCS Subsystems - Shutdown 3.5.3 YES-3 3/4.5.5 Refueling Water Storage Tank 3.5.5 Refueling Water Storage Tank 3.5.4 YES-3 3/4.5.6 Seal Injection Flow 3.5.6 Seal Injection Flow 3.5.5 YES-2 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 12 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.6 CONTAINMENT SYSTEMS 3.6 3/4.6.1 Primary Containment 3.6.1.1 Containment Integrity 3.6.1 3.6.2 YES-3 3.6.1.3 Containment Air Locks 3.6.2 YES-3 3.6.1.4 Internal Pressure 3.6.4 YES-2 3.6.1.5 Air Temperature 3.6.5 YES-2 3.6.1.6 Containment Vessel Structural Integrity 3.6.1 YES-3 Containment vessel structural integrity is being retained as a Surveillance Requirement (SR 3.6.1.1) in ITS 3.6.1. 3.6.1.7 Shield Building Structural Integrity 3.6.7 YES-3 3.6.1.8 Emergency Gas Treatment System-EGTS-Clean Up Subsystems 3.6.10 5.5.9 YES-3 3/4.6.2 Depressurization and Cooling Systems 3.6.2.1 Containment Spray SubSystems 3.6.6 YES-3 3.6.2.2 Lower Containment Vent Coolers Relocated No See Appendix A, Page 2. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 13 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.6.3 Containment Isolation Valves 3.6.3 YES-3 3.6.3 Containment Isolation Valves 3.6.3 YES-3 3/4.6.4 Combustible Gas Control 3.6.4.3 Hydrogen Mitigation System 3.6.8 YES-4 3/4.6.5 Ice Condenser 3.6.5.1 Ice Bed 3.6.12 YES-3 3.6.5.3 Ice Condenser Doors 3.6.13 YES-3 3.6.5.5 Divider Barrier Personnel Access Doors and Equipment hatches 3.6.14 YES-3 3.6.5.6 Containment Air Return Fans 3.6.11 YES-3 3.6.5.7 Floor Drains 3.6.15 YES-3 3.6.5.8 Refueling Canal Drains 3.6.15 YES-3 3.6.5.9 Divider Barrier Seals 3.6.14 YES-3 3/4.6.6 Vacuum Relief Lines 3.6.6 Vacuum Relief lines 3.6.9 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 14 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.7 PLANT SYSTEMS 3.7 3/4.7.1 Turbine Cycle 3.7.1.1 Safety Valves 3.7.1 YES-3 3.7.1.2 Auxiliary Feedwater (AFW) System 3.7.5 YES-3 3.7.1.3 Condensate Storage System 3.7.6 YES-2, 3 3.7.1.4 Activity 3.7.16 YES-2 3.7.1.5 Main Steam Line Isolation Valves 3.7.2 YES-3 3.7.1.6 Main Feedwater Isolation, Regulation and Bypass Valves 3.7.3 YES-3 3/4.7.3 Component Cooling Water System 3.7.3 Component Cooling Water System 3.7.7 YES-3 3/4.7.4 Service Water System 3.7.4 Service Water System 3.7.8 YES-3 3/4.7.5 Ultimate Heat Sink 3.7.5 Ultimate Heat Sink 3.7.9 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 15 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.7.7 Control Room Emergency Ventilation System 3.7.7 Control Room Emergency Ventilation System 3.7.10 5.5.9 YES-3 3/4.7.8 Auxiliary Building Gas Treatment System 3.7.8 Auxiliary Building Gas Treatment System 3.7.12 5.5.9 YES-3 3/4.7.13 Spent Fuel Pool Minimum Boron Concentration 3.7.13 Spent Fuel Pool Minimum Boron Concentration 3.7.14 YES-2 3/4.7.14 Cask Pit Pool Minimum Boron Concentration 3.7.14 Cask Pit Pool Minimum Boron Concentration 3.7.17 YES-2 3/4.7.15 Control Room Air Conditioning System 3.7.15 Control Room Air Conditioning System 3.7.11 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 16 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.8 ELECTRICAL POWER SYSTEM 3.8 3/4.8.1 A.C. Sources 3.8.1.1 Operating 3.8.1 3.8.3 3.8.4 3.8.6 3.8.9 YES-3 3.8.1.2 Shutdown 3.8.2 3.8.3 3.8.5 3.8.6 3.8.10 YES-3 3/4.8.2 Onsite Power Distribution Systems 3.8.2.1 A.C. Distribution - Operating 3.8.7 3.8.9 YES-3 3.8.2.2 A.C. Distribution - Shutdown 3.8.8 3.8.10 YES-3 3.8.2.3 D.C. Distribution - Operating 3.8.4 3.8.6 3.8.9 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 17 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3.8.2.4 D.C. Distribution - Shutdown 3.8.5 3.8.6 3.8.10 YES-33/4.9 REFUELING OPERATIONS 3.9 3/4.9.1 Boron Concentration 3.9.1 Boron Concentration 3.9.1 3.9.2 YES-23/4.9.2 Instrumentation 3.9.2 Instrumentation 3.9.3 YES-3 3/4.9.3 Decay Time 3/4.9.3 Decay Time Deleted NO See technical change discussion in Enclosure 2, Volume 14, Discussion of Changes for CTS 3/4.9.3. 3/4.9.4 Containment Building Penetrations 3.9.4 Containment Building Penetrations 3.9.4 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 18 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.9.8 Residual Heat Removal and Coolant Circulation 3.9.8.1 All Water Levels 3.9.5 3.9.6 YES-4 3.9.8.2 Low Water Level 3.9.6 YES-4 3/4.9.9 Containment Ventilation Isolation System 3.9.9 Containment Ventilation Isolation System 3.9.4 YES-3 3/4.9.10 Water Level - Reactor Vessel 3.9.10 Water Level - Reactor Vessel 3.9.7 YES-2 3/4.9.11 Storage Pool Water Level 3.9.11 Storage Pool Water Level 3.7.13 YES-2, 3 3/4.9.12 Auxiliary Building Gas Treatment System 3.9.12 Auxiliary Building Gas Treatment System 3.7.12 5.5.9 YES-3 SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 19 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 Shutdown Margin 3.10.1 Shutdown Margin Deleted NO See technical change discussion in Enclosure 2, Volume 6, Discussion of Changes for CTS 3/4.10.1. 3/4.10.2 Group Height, Insertion and Power Distribution Limits 3/4.10.2 Group Height, Insertion and Power Distribution Limits Deleted NO See technical change discussion in Enclosure 2, Volume 6, Discussion of Changes for CTS 3/4.10.2. 3/4.10.3 Physics Tests 3.10.3 Physics Tests 3.1.8 YES Although this Specification does not meet any Technical Specification selection criteria, it has been retained to provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 20 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3/4.10.4 Reactor Coolant Loops 3.10.4 Reactor Coolant Loops Deleted NO See technical change discussion in Enclosure 2, Volume 6, Discussion of Changes for CTS 3/4.10.4. 3/4.11 RADIOACTIVE EFFLUENTS NA 3/4.11.1 Liquid Effluents 3.11.1.4 Liquid Holdup Tanks 5.5.10 YES Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W. T. Russell to the industry ITS Chairpersons, dated October 25, 1993. 3/4.11.2 Gaseous Effluents 3.11.2.5 Explosive Gas Mixture 5.5.10 YES Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W. T. Russell to the industry ITS Chairpersons, dated October 25, 1993. SUMMARY DISPOSITION MATRIX FOR SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 1 The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. Page 21 of 21 CURRENT TS (CTS) NUMBER CURRENT TITLE NEW TS (ITS) NUMBER RETAINED/ CRITERION FOR INCLUSION NOTES(1) 3.11.2.6 Gas Decay Tanks 5.5.10 YES Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W. T. Russell to the industry ITS Chairpersons, dated October 25, 1993. 5.0 DESIGN FEATURES 3.7.15 4.0 YES-2 YES Application of Technical Specification selection criteria is not appropriate. However, specific portions of Design Features will be included in Technical Specifications as required by 10 CFR 50.36. 6.0 ADMINISTRATIVE CONTROLS 5.0 YES Application of Technical Specification selection criteria is not appropriate. However, specific portions of Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36. APPENDIX A JUSTIFICATION FOR SPECIFICAITON RELOCATION Appendix A - Justification For Specification Relocation Page 1 of 2 3.3.3.10, Explosive Gas Monitoring Instrumentation DISCUSSION: CTS 3.3.3.10 provides the requirements for the explosive gas monitoring instrumentation. This Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the gaseous waste processing system is adequately monitored to ensure that the concentration is maintained below the flammability limit. COMPARISON TO SCREENING CRITERIA: 1. Explosive gas monitoring instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA. 2. Explosive gas monitoring instrumentation is not used to indicate the status of, or monitor a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient. In addition, excessive system oxygen is not an indication of a DBA or transient.

3. Explosive gas monitoring instrumentation is not part of a primary success path in the mitigation of a DBA or transient. In addition, excessive oxygen discharge is not part of a primary success path in mitigating a DBA or transient.

4. As discussed in Section 4.0 (Appendix A, page A-69) and summarized in Table 1 of WCAP-11618, the loss of the explosive gas monitoring instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases. TVA has reviewed this evaluation, considers it applicable to Sequoyah Nuclear Plant (SQN) Units 1 and 2, and concurs with the assessment.

CONCLUSION: Since the screening criteria have not been satisfied, Explosive Gas Monitoring Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside Technical Specifications.

Appendix A - Justification For Specification Relocation Page 2 of 2 3.6.2.2, Lower Containment Vent Coolers DISCUSSION: CTS 3.6.2.2 provides requirements on the Lower Containment Vent Coolers. The Lower Containment Vent Coolers are designed to maintain an acceptable temperature within the lower containment compartments for the protection of equipment and controls during normal reactor operation and normal shutdown. COMPARISON TO SCREENING CRITERIA: 1. The Lower Containment Vent Coolers are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.

2. The Lower Containment Vent Coolers are not a process variable, design feature, or operating restriction that is in an initial condition of a DBA or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. 3. The Lower Containment Vent Coolers are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. The Lower Containment Vent Coolers were found to be non-significant risk contributor to core damage frequency and offsite releases. Tennessee Valley Authority (TVA) has performed a plant-specific analysis to ensure that the Lower Containment Vent Coolers do not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to be important to public health and safety. CONCLUSION: Since the screening criteria have not been satisfied, the Lower Containment Vent Coolers may be relocated to other plant controlled documents outside Technical Specifications.

ENCLOSURE 2 VOLUME 3 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 1.0 USE AND APPLICATION Revision 0 LIST OF ATTACHMENTS 1. ITS Chapter 1.0 - USE AND APPLICATION ATTACHMENT 1 ITS 1.0, USE AND APPLICATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS Chapter 1.01.0 DEFINITIONS DEFINED TERMS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector. BYPASS LEAKAGE PATH 1.3 A BYPASS LEAKAGE PATH is a potential path for leakage to escape from both the primary containment and annulus pressure boundary. Only one type of BYPASS LEAKAGE PATH is recognized:

a. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING are those paths that would potentially allow leakage from the primary containment to circumvent the annulus secondary containment enclosure and escape directly to the Auxiliary Building secondary containment enclosure. CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

May 18, 1988 SEQUOYAH - UNIT 1 1-1 Amendment No. 12, 71 A01A01A01A03A01A011.0 USE AND APPLICATION Definitions 1.1 and Bases (AFD) 1 A01AFDINSERT 2thatmeans of Required Actions to be taken that , by observation, to NOTE ACTIONS AXIAL FLUX DIFFERENCE (AFD) CHANNEL CALIBRATION CHANNEL CHECK Page 1 of 37 Swithin specified Completion Times A02INSERT 1 within , See ITS 3.6.3 Term Definition Chapter 1.0 Insert Page 1-1 Page 2 of 37INSERT 1 ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

INSERT 2 all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. A02A03 ITS Chapter 1.0DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be: a.Analog channels - the injection of a simulated signal into the channel as close to thesensor as practicable to verify OPERABILITY including alarm and/or trip functions.b.Bistable channels - the injection of a simulated signal into the sensor to verifyOPERABILITY including alarm and/or trip functions.c.Digital channels - the injection of a simulated signal into the channel as close to the sensorinput to the process racks as practicable to verify OPERABILITY including alarm and/ortrip functions. CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: a.All penetrations required to be closed during accident conditions are either:1)Capable of being closed by an OPERABLE containment automatic isolation valvesystem, or 2)Closed by manual valves, blind flanges, or deactivated automatic valves secured intheir closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3. b.All equipment hatches are closed and sealed.c.Each air lock is in compliance with the requirements of Specification 3.6.1.3,d.The containment leakage rates are within the limits of Specification 4.6.1.1.c,e.The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings)is OPERABLE, andf.Secondary containment bypass leakage is within the limits of Specification 3.6.3.CONTROLLED LEAKAGE 1.8 This definition has been deleted. CORE ALTERATION 1.9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. CORE OPERATING LIMIT REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these operating limits is addressed in individual specifications. April 13, 2009 SEQUOYAH - UNIT 1 1-2 Amendment No. 12, 71, 130, 141, 155 176, 201, 203, 259, 323 A04A04L01A05A06A07A06A01OPERATIONAL 5.6.3. Plantparameteror actual(COT)COT INSERT 3 A01(COLR)CHANNEL OPERATIONAL TEST A04S CORE OPERATING LIMITS REPORT cycle specific parameter Page 3 of 37 Chapter 1.0 Insert Page 1-2 INSERT 3 of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps. A04Page 4 of 37 ITS Chapter 1.0DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." _ E - AVERAGE DISINTEGRATION ENERGY _ 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC. FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. GASEOUS RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be: a. Leakage, such as that from pump seals or valve packing (except Reactor Coolant Pump Seal Water Injection or Leakoff), that is captured and conducted to collection systems or a sump or collecting tank, or

August 4, 2000 SEQUOYAH - UNIT 1 1-3 Amendment No. 12, 71, 155, 251, 259 A01A02A06A01LA02A06A08that INSERT 4 A01a. Identified LEAKAGE1. (RCP) A01ESFAEC, 1962, DOSE EQUIVALENT I-131 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE Page 5 of 37 the, A01when inhaled as the combined activities of iodine isotopes determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from Chapter 1.0 Insert Page 1-3 INSERT 4 DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."Page 6 of 37TSTF-490 ITS Chapter 1.0b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage). MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s). OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1. PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

February 23, 2006 SEQUOYAH - UNIT 1 1-4 Amendment No. 12, 71, 148, 155, 169, 174, 178, 281, 306 2. 3. (RCS)A01INSERT 5 A02A07specified safetyorand A01A09A10INSERT 6 ,-1 with fuel in the reactor vessel A11A01. These tests are:¶ a. c.Nuclear Regulatoryb.A01safety LEAKAGE Page 7 of 37 OPERABLE - OPERABILITY MODE PHYSICS TESTS ,, and , See ITS 5.5 , Initial Tests and Operations, Chapter 1.0 Insert Page 1-4 INSERT 5 MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. INSERT 6

, and reactor vessel head closure bolt tensioning

A02A11Page 8 of 37 ITS Chapter 1.0PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.23 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the LTOP arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.15. PROCESS CONTROL PROGRAM (PCP) 1.24 DELETED PURGE - PURGING 1.25 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RATED THERMAL POWER (RTP) 1.27 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3455 MWt. REACTOR TRIP SYSTEM (RTS) RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC. REPORTABLE EVENT 1.29 DELETED

February 23, 2006 SEQUOYAH - UNIT 1 1-5 Amendment No. 12, 71, 141, 148, 155, 201, 233, 251, 275, 276, 294, 297, 306 c.an RCSA08A01low temperature overpressure protectionA01A07A06(QPTR)QPTRA01A01thatA01A07LEAKAGE PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) REACTOR TRIP SYSTEM (RTS) RESPONSE TIME Page 9 of 37 RATED THERMAL POWER (RTP) QUADRANT POWER TILT RATIO (QPTR) ,5.6.4the RTS ITS Chapter 1.0SHIELD BUILDING INTEGRITY 1.30 SHIELD BUILDING INTEGRITY shall exist when: a.The door in each access opening is closed except when the access opening is being usedfor normal transit entry and exit.b.The emergency gas treatment system is OPERABLE.c.The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings)is OPERABLE.SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY 1.32 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. SOLIDIFICATION 1.33 Deleted SOURCE CHECK 1.34 Deleted STAGGERED TEST BASIS 1.35 A STAGGERED TEST BASIS shall consist of: a.A test schedule for n systems, subsystems, trains or other designated componentsobtained by dividing the specified test interval into n equal subintervals,b.The testing of one system, subsystem, train or other designated component at thebeginning of each subinterval. THERMAL POWER 1.36 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. August 2, 2006 SEQUOYAH - UNIT 1 1-6 Amendment No. 12, 71, 48, 155, 294, 297, 309 (SDM) SDM: a. RCCARCCAs INSERT 7A01A12A06A07A07INSERT 9 A13A02A01, SHUTDOWN MARGIN (SDM) STAGGERED TEST BASIS THERMAL POWER Page 10 of 37See ITS 3.6.3 See ITS 3.6.13 See ITS 3.6.1 controlINSERT 8 See ITS 3.6.3 3.6.13 3.6.1 Chapter 1.0 Insert Page 1-6 INSERT 7 With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level. INSERT 8 SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

INSERT 9

the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. A02A12A13Page 11 of 37 ITS Chapter 1.0UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage (except reactor coolant pump seal water injection or leakoff) that is not IDENTIFIED LEAKAGE. UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commerical, institutional, and/or recreational purposes. VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.40 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

November 9, 2004 SEQUOYAH - UNIT 1 1-7 Amendment No. 12, 71, 155, 259, 294, 297 , and b. RCP All A08A01A06A06A06INSERT 10A02LEAKAGE Page 12 of 37 Chapter 1.0 Insert Page 1-7 INSERT 10 TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating OPERATIONAL TEST device and verifying the OPERABILITY of all devices in the (TADOT) channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. A02Page 13 of 37 ITS Chapter 1.0TABLE 1.1 OPERATIONAL MODES MODE REACTIVITY CONDITION, Keff % RATED THERMAL POWER* AVERAGE COOLANT TEMPERATURE 1. POWER OPERATION 0.99 >5% 350oF 2. STARTUP 0.99 5% 350oF 3. HOT STANDBY <0.99 0 350oF

4. HOT SHUTDOWN <0.99 0 350oF >Tavg >200oF

5. COLD SHUTDOWN <0.99 0 200oF 6. REFUELING** 0.95 0 140oF

_________________ *Excluding decay heat. **Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

June 1, 1995 SEQUOYAH - UNIT 1 1-8 Amendment No. 71, 201 TITLE -1 (page 1 of 1) (b) A01A01A01A14A11(a) TABLE 1.1-1 LA01Page 14 of 37A01(a) REACTOR(0F)(c) NA (b) All reactor vessel head closure bolts fully tensioned. (c) One or more reactor ITS Chapter 1.0TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release. N.A. Not applicable.

May 18, 1988 SEQUOYAH - UNIT 1 1-9 Amendment No. 12, 71 LA02Add proposed ITS Sections 1.2 - Logical Connectors 1.3 - Completion Times 1.4 - Frequency A15Page 15 of 37 ITS Chapter 1.03/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg Greater Than 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

____________________ *See Special Test Exception 3.10.1 November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-1 Amendment No. 172 Page 16 of 37A12SHUTDOWN MARGIN (SDM) See ITS 3.1.1 See ITS 3.1.4 See ITS 3.1.6 See ITS 3.1.1 ITS Chapter 1.0REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-3 Amendment No. 12, 172 Page 17 of 37 SHUTDOWN MARGIN (SDM) A12See ITS 3.1.1 See ITS 3.1.4 See ITS 3.1.1 Chapter 1.0 ITS 1.0 DEFINITIONS DEFINED TERMS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions. AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detectors. BYPASS LEAKAGE PATH 1.3 A BYPASS LEAKAGE PATH is a potential path for leakage to escape from both the primary containment and annulus pressure boundary. Only one type of BYPASS LEAKAGE PATH is recognized: a. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING are those paths that would potentially allow leakage from the primary containment to circumvent the annulus secondary containment enclosure and escape directly to the auxiliary building secondary containment enclosure. CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

May 18, 1988 SEQUOYAH - UNIT 2 1-1 Amendment No. 63 NOTE ACTIONS AXIAL FLUX DIFFERENCE (AFD) CHANNEL CALIBRATION CHANNEL CHECK and Bases A01that Required Actions to be taken A01A01A02INSERT 1 (AFD) AFD A01that means of INSERT 2A03A01A01, by observation, to A01Page 18 of 37 A011.0 USE AND APPLICATION Definitions 1.1 1 within specified Completion Times See ITS 3.6.3 ,within Term Definition S Chapter 1.0 Insert Page 1-1 Page 19 of 37INSERT 1 ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

INSERT 2 all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. A02A03 Chapter 1.0 ITS DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be: a.Analog channels - the injection of a simulated signal into the channel as close to the sensoras practicable to verify OPERABILITY including alarm and/or trip functions.b.Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITYincluding alarm and/or trip functions.c.Digital channels - the injection of a simulated signal into the channel as close to the sensorinput to the process racks as practicable to verify OPERABILITY including alarm and/or tripfunctions.CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: a.All penetrations required to be closed during accident conditions are either:1)Capable of being closed by an OPERABLE containment automatic isolation valvesystem, or 2)Closed by manual valves, blind flanges, or deactivated automatic valves secured intheir closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3. b.All equipment hatches are closed and sealed.c.Each air lock is in compliance with the requirements of Specification 3.6.1.3,d.The containment leakage rates are within the limits of Specification 4.6.1.1.c,e.The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) isOPERABLE, and f.Secondary containment bypass leakage is within the limits of Specification 3.6.3.CONTROLLED LEAKAGE 1.8 This definition has been deleted. CORE ALTERATION 1.9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.14. Unit operation within these operating limits is addressed in individual specifications. April 13, 2009 SEQUOYAH - UNIT 2 1-2 Amendment Nos. 63, 117, 132, 146, 167, 191, 193, 250, 315 CHANNEL OPERATIONAL TEST OPERATIONAL (COT)COT or actual INSERT 3 A04A01L01A04A05A04A06A07A06CORE OPERATING LIMITS REPORT (COLR) A01Page 20 of 37 5.6.3. Plant parameter cycle specific parameter Chapter 1.0 Insert Page 1-2 INSERT 3 of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps. A04Page 21 of 37 Chapter 1.0 ITS DEFINITIONS DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." _ E - AVERAGE DISINTEGRATION ENERGY _ 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC. FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. GASEOUS RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

February 29, 2000 SEQUOYAH - UNIT 2 1-3 Amendment Nos. 63, 146, 242 that A01AEC, 1962, DOSE EQUIVALENT I-131 A02A06ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME ESF A01A01LA02A06Page 22 of 37 , determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from when inhaled as the combined activities of iodine isotopes INSERT 4 Chapter 1.0 Insert Page 1-3 INSERT 4 DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."Page 23 of 37TSTF-490 Chapter 1.0 ITS DEFINITIONS IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage, such as that from pump seals or valve packing (except reactor coolant pump seal injection or leakoff) that is captured and conducted to collection systems or a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

May 22, 2007 SEQUOYAH - UNIT 2 1-4 Amendment Nos. 63, 134, 146, 159, 165, 169, 250, 272, 305 LEAKAGE a. Identified LEAKAGE (RCP) A01A082. 3. A01(RCS) INSERT 5 A02A07See ITS 5.5 OPERABLE - OPERABILITY , safety or and, and specified safety A01A09A10Page 24 of 37 1. water , Chapter 1.0 Insert Page 1-4 INSERT 5 MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. A02Page 25 of 37 Chapter 1.0 ITS DEFINITIONS OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.23 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the LTOP arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.15. PROCESS CONTROL PROGRAM (PCP) 1.24 DELETED PURGE - PURGING 1.25 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, which-ever is greater.

May 22, 2007 SEQUOYAH - UNIT 2 1-5 Amendment No. 63, 134, 146, 191, 223, 284, 305 MODE , INSERT 6-1 with fuel in the reactor vessel A11A01PHYSICS TESTS , Initial Tests and Operations,Nuclear Regulatory b.c.. These tests are: a. A01LEAKAGE an RCSc.A08A01low temperature overpressure protection5.6.4 A01A07A06PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (QPTR)QPTR QUADRANT POWER TILT RATIO (QPTR) A01Page 26 of 37 , Chapter 1.0 Insert Page 1-5 INSERT 6 , and reactor vessel head closure bolt tensioning A11Page 27 of 37 Chapter 1.0 ITS DEFINITIONS RATED THERMAL POWER (RTP) 1.27 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3455 MWt. REACTOR TRIP SYSTEM (RTS) RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC. REPORTABLE EVENT 1.29 DELETED SHIELD BUILDING INTEGRITY 1.30 SHIELD BUILDING INTEGRITY shall exist when: a.The door in each access opening is closed except when the access opening is beingused for normal transit entry and exit.b.The emergency gas treatment system is OPERABLE.c.The sealing mechanism associated with each penetration (e.g., welds, bellows orO-rings) is OPERABLE.SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY 1.32 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. August 2, 2006 SEQUOYAH - UNIT 2 1-6 Amendment No. 63, 132, 146, 242, 264, 267, 284, 298 RATED THERMAL POWER (RTP) A01RTS that the REACTOR TRIP SYSTEM (RTS) RESPONSE TIME A01A07Page 28 of 37 See ITS 3.6.3 See ITS 3.6.13 See ITS 3.6.1 See ITS 3.6.3 3.6.13 3.6.1 SHUTDOWN MARGIN (SDM) (SDM) SDM : a. RCCA INSERT 7 RCCAs , A01A12A06control Chapter 1.0 Insert Page 1-6 INSERT 7 With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level. A12Page 29 of 37 Chapter 1.0 ITS DEFINITIONS SOLIDIFICATION 1.33 Deleted.

SOURCE CHECK 1.34 Deleted. STAGGERED TEST BASIS 1.35 A STAGGERED TEST BASIS shall consist of: a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals,

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER 1.36 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage (except reactor coolant pump seal water injection or leakoff) that is not IDENTIFIED LEAKAGE. UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

September 15, 2004 SEQUOYAH - UNIT 2 1-7 Amendment Nos. 63, 134, 146, 250, 284 A07A07INSERT 8 INSERT 9 A02A13STAGGERED TEST BASIS THERMAL POWER A01LEAKAGE b. All , and RCP A08A01A06Page 30 of 37 Chapter 1.0 Insert Page 1-7 INSERT 8 SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. INSERT 9

the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. A02A13Page 31 of 37 Chapter 1.0 ITS DEFINITIONS VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.40 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

September 15, 2004 SEQUOYAH - UNIT 2 1-8 Amendment Nos. 63, 146, 284 A06A06INSERT 10 A02Page 32 of 37 Chapter 1.0 Insert Page 1-8 INSERT 10 TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating OPERATIONAL TEST device and verifying the OPERABILITY of all devices in the (TADOT) channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. A02Page 33 of 37 Chapter 1.0 ITS TABLE 1.1 OPERATIONAL MODES MODE REACTIVITY CONDITION, Keff % RATED THERMAL POWER* AVERAGE COOLANT TEMPERATURE

1. POWER OPERATION 0.99 > 5% 350F

2. STARTUP 0.99 5% 350F

3. HOT STANDBY < 0.99 0 350F

4. HOT SHUTDOWN < 0.99 0 350F > Tavg > 200F

5. COLD SHUTDOWN < 0.99 0 200F 6. REFUELING** 0.95 0 140F

Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

May 18, 1988 SEQUOYAH - UNIT 2 1-9 Amendment No. 63 (b) All reactor vessel head closure bolts fully tensioned. (a) (c) One or more reactor A11A01NA LA01A14TABLE 1.1-1 -1 (page 1 of 1) TITLE (a) REACTOR(0F)A01A01A01(b) (c) Page 34 of 37 Chapter 1.0 ITS TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. M At least once per 31 days Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release. N.A. Not applicable.

May 18, 1988 SEQUOYAH - UNIT 2 1-10 Amendment No. 63 Add proposed ITS Sections 1.2 - Logical Connectors 1.3 - Completion Times 1.4 - Frequency A15LA02Page 35 of 37 Chapter 1.0 ITS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg 200F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2, with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-1 Amendment No. 163 Page 36 of 37See ITS 3.1.1 See ITS 3.1.6 See ITS 3.1.1 See ITS 3.1.4 A12 Chapter 1.0 ITS REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and

6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-3 Amendment No. 163 Page 37 of 37See ITS 3.1.4 A12See ITS 3.1.1 See ITS 3.1.1 DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 1 of 11 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN), Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional approved Technical Specification Task Force (TSTF) travelers included in the submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 ITS Section 1.1 provides definitions of ACTUATION LOGIC TEST, MASTER RELAY TEST, SLAVE RELAY TEST, DOSE EQUIVALENT XENON XE-133, and TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT). These terms are used as defined terms in the ITS but do not appear in the CTS. This changes the CTS by adding new definitions This change is acceptable because these new defined terms, of themselves, do not impose any new requirements or alter existing requirements. Any technical changes due to the addition of these defined terms are addressed in the discussion of changes (DOCs) for the sections of the Technical Specifications in which the terms are used. These changes are designated as administrative as they add defined terms that do not involve a technical change to the Technical Specifications. A03 CTS 1.4 defines a CHANNEL CALIBRATION as "the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated." ITS defines a CHANNEL CALIBRATION as "the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps." This results in a number of changes to the CTS.

The CTS definition states, "The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions." The ITS states, "The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY."

DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 2 of 11 This change is acceptable because the statements are equivalent in that both require that all needed portions of the channel be tested. The ITS definition reflects the CTS understanding that the CHANNEL CALIBRATION includes only those portions of the channel needed to perform the safety function.

The CTS states that the CHANNEL CALIBRATION "shall include the CHANNEL FUNCTIONAL TEST." The ITS does not include this statement.

This change is acceptable because the eliminated CTS statement does not add any requirements. In both the CTS and the ITS, performance of a single test that fully meets the requirements of another test can be credited as satisfying that other test.

The ITS adds the statement, "Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel." The purpose of a CHANNEL CALIBRATION is to adjust the channel output so that the channel responds within the necessary range and accuracy to known values of the parameters that the channel monitors. This change is acceptable because resistance temperature detectors and thermocouples are designed such that they have a fixed input/output response, which cannot be adjusted or changed once installed. Calibration of a channel containing an RTD or thermocouple is performed by applying the RTD or thermocouple fixed input/output relationship to the remainder of the channel, and making the necessary adjustments to the adjustable devices in the remainder of the channel to obtain the necessary output range and accuracy. Therefore, unlike other sensors, an RTD or thermocouple is not actually calibrated. The ITS CHANNEL CALIBRATION allowance for channels containing RTDs and thermocouples is consistent with the CTS calibration practices of these channels. This information is included in the ITS to avoid confusion, but does not change the current CHANNEL CALIBRATION practices for these types of channels.

These changes are designated as administrative because they do not result in a technical change to the Technical Specifications. A04 CTS Section 1.0 defines CHANNEL FUNCTIONAL TEST as: "a. Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions; b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions; c. Digital channels - the injection of a simulated signal into the channel as close to the sensor input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions ." ITS Section 1.1 renames and combines the CTS definition to CHANNEL OPERATIONAL TEST (COT), and defines it as "the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 3 of 11 required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps." This changes the CTS by stating that the COT shall include adjustments, as necessary, of the devices in the channel so that the setpoints are within the required range and accuracy, changes the CTS by combining the type of devices contained in the definition, and states that the test may be performed by means of any series of sequential, overlapping, or total channel steps. The addition of use of an actual signal is discussed in DOC L01. The CTS definition states that the CHANNEL FUNCTIONAL TEST shall verify that the channel is OPERABLE "including alarm and/or trip functions." Similarly, the ITS requirement states that the COT shall verify OPERABILITY of "all devices in the channel required for channel OPERABILITY." This change is acceptable because the statements are equivalent in that both require verification of channel OPERABILITY. The CTS and the ITS use different examples of what is included in a channel, but this does not change the intent of the requirement. The ITS use of the phrase "all devices in the channel required for channel OPERABILITY," reflects the CTS understanding that the test includes only those portions of the channel needed to perform the specified safety function(s). The ITS requirement states "The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy." This change is acceptable because it clarifies that adjustments performed during a COT do not invalidate the test. This is consistent with the current implementation of the CHANNEL FUNCTIONAL TEST and does not result in a technical change to the Technical Specifications. The ITS states "The COT may be performed by means of any series of sequential, overlapping, or total channel steps." This change is acceptable because it states current Industry practice and is consistent with the current implementation of the CHANNEL FUNCTIONAL TEST. Therefore, this change does not result in a technical change to the Technical Specifications. CTS Section 1.0 defines CHANNEL FUNCTIONAL TEST for analog channels and digital channels. The ITS definition combines theses definitions. This change is acceptable because it states current Industry practice and is consistent with the current implementation of the CHANNEL FUNCTIONAL TEST. This conclusion was confirmed when the NRC issued SQN Unit 1/Unit 2 License Amendment 140/132 (ADAMS Accession Nos. ML013310103 / ML013330076) concluding that the addition of the definition DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 4 of 11 to the CHANNEL FUNCTIONAL TEST for digital channels was consistent with the existing channel functional test definition and therefore acceptable. These changes are designated as administrative because they do not result in a technical change to the Technical Specifications. A05 CTS Section 1.0 includes a CHANNEL FUNCTIONAL TEST definition for bistable channels. The definition of CHANNEL FUNCTIONAL TEST for bistable channels requires "the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions." However, this CTS definition is essentially duplicative of the TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) definition. ITS Section 1.1 does not include this definition, since the requirements for bistable channels are covered by the TADOT definition. This change is acceptable because the TADOT definition adequately covers bistable channels, and does not impose any new requirements or alter any existing requirements. This change is categorized as administrative because the bistable portion of the definition is duplicative of the TADOT definition. A06 CTS Section 1.0 includes the following definitions: CONTAINMENT INTEGRITYGASEOUS RADWASTE TREATMENT SYSTEMPURGE - PURGINGSITE BOUNDARYUNRESTRICTED AREAVENTILATION EXHAUST TREATMENT SYSTEMVENTINGE - AVERAGE DISINTEGRATION ENERGYCORE ALTERATIONThe ITS does not use this terminology and ITS Section 1.1 does not contain these definitions. These changes are acceptable because the terms are not used as defined terms in the ITS. Discussions of any technical changes related to the deletion of these terms are included in the DOCs for the CTS sections in which the terms are used. These changes are designated as administrative because they eliminate defined terms that are no longer used. A07 CTS Section 1.0 shows the following definitions as being deleted: CONTROLLED LEAKAGEMEMBER(S) OF THE PUBLICPROCESS CONTROL PROGRAM (PCP)REPORTABLE EVENTSOLIDIFICATIONSOURCE CHECK DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 5 of 11 The ITS does not use this terminology and ITS Section 1.1 does not contain these definitions. These changes are acceptable because the terms are not used as defined terms in the ITS. Previous license amendments have deleted these definitions. This change removes the placeholder showing these definitions as deleted. These changes are designated as administrative because they eliminate deleted defined terms that are no longer used. A08 CTS Section 1.0 provides definitions for IDENTIFIED LEAKAGE, PRESSURE BOUNDARY LEAKAGE, and UNIDENTIFIED LEAKAGE. ITS Section 1.1 includes these requirements in one definition called LEAKAGE (which includes three categories: identified LEAKAGE, unidentified LEAKAGE, and pressure boundary LEAKAGE). This changes the CTS by incorporating the definitions into the ITS LEAKAGE definition with no technical changes. This change is acceptable because it results in no technical changes to the Technical Specifications. This change is designated an administrative change in that it rearranges existing definitions, with no change in intent. A09 The CTS Section 1.0 definition of OPERABLE - OPERABILITY requires a system, subsystem, train, component, or device to be capable of performing its "specified function(s)" and all necessary support systems to also be capable of performing their "function(s)." The ITS Section 1.1 definition of OPERABLE - OPERABILITY requires the system, subsystem, train, component, or device to be capable of performing the "specified safety function(s)," and requires all necessary support systems that are required for the system, subsystem, train, component, or device to perform its "specified safety function(s)" to also be capable of performing their related support functions. This changes the CTS by altering the requirement to be able to perform "functions" to a requirement to be able to perform "safety functions." The purpose of the CTS and ITS definitions of OPERABLE - OPERABILITY are to ensure that the safety analysis assumptions regarding equipment and variables are valid. This change is acceptable because the intent of both the CTS and ITS definitions is to address the safety function(s) assumed in the accident analysis and not encompass other non-safety functions a system may also perform. These non-safety functions are not assumed in the safety analysis and are not needed in order to protect the public health and safety. This change is consistent with the current interpretation and use of the terms OPERABLE and OPERABILITY. This change is designated as administrative as it does not change the current use and application of the Technical Specifications. A10 The CTS Section 1.0 definition of OPERABLE - OPERABILITY requires that all necessary normal and emergency electrical power sources be available for the system, subsystem, train, component, or device to be OPERABLE. The ITS Section 1.1 definition of OPERABLE - OPERABILITY will replace the phrase "normal and emergency electrical power sources" with "normal or emergency electrical power." This changes the CTS definition of OPERABLE - OPERABILITY by allowing a device to be considered OPERABLE with either normal or emergency power available. DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 6 of 11 The OPERABILITY requirements for normal and emergency power sources are addressed in CTS 3.0.5. These requirements allow only the normal or the emergency electrical power source to be OPERABLE, provided its redundant system(s), subsystem(s), train(s), component(s), and device(s) (redundant to the systems, subsystems, trains, components, and devices with an inoperable power source) are OPERABLE. This effectively changes the current "and" to an "or." The existing CTS 3.0.5 requirements are incorporated into ITS 3.8.1 ACTIONS for when a normal (offsite) or emergency (diesel generator) power source is inoperable. Therefore, the ITS definition now uses the word "or" instead of the current word "and." In ITS 3.8.1, new times are provided to perform the determination of OPERABILITY of the redundant systems. This change is discussed in the Discussion of Changes (DOCs) for ITS 3.8.1. This change is designated administrative since the ITS definition is effectively the same as the CTS definition. A11 CTS Section 1.0 and Table 1.1, "OPERATIONAL MODES," provide a description of the MODES. CTS Section 1.0 and Table 1.1 contains Note ** that states, "Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed." ITS Section 1.1 and Table 1.1-1, "MODES," changes the CTS MODE definitions in the following ways: The CTS Table 1.1 Note ** condition "fuel in the vessel" is moved to the ITS MODE definition. This change is acceptable because it moves information within the Technical Specifications with no change in intent. Each MODE in the Table includes fuel in the vessel. CTS Table 1.1, Note ** in part states, "-with the vessel head closure bolts less than fully tensioned or with the head removed." ITS splits this portion of the Note into two Notes, Notes (b), and (c). ITS Note (b) states, "All reactor vessel head closure bolts fully tensioned," while Note (c) states, "One or more reactor vessel head closure bolts less than fully tensioned." This change simplifies what CTS is stating by clearly defining when the reactor is in a refueling condition instead of a shutdown condition. This change is acceptable because the revised phrase is consistent with the current interpretation and usage. MODE 6 is currently declared when the first vessel head closure bolt is detensioned. This change also eliminates a redundant phrase. The reactor vessel head cannot be removed unless the reactor vessel head closure bolts are unbolted and they cannot be unbolted unless they are detensioned. Since "reactor vessel head unbolted" is already specified in the CTS Note, including "or removed" is unnecessary. ITS Table 1.1-1 contains a new Note b, which applies to MODES 4 and 5. Note b states "All reactor vessel head closure bolts fully tensioned." This Note is the opposite of CTS Note ** and ITS Table 1.1-1 Note (c). DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 7 of 11 This change is acceptable because it avoids a conflict between the definition of MODE 6 and the other MODES should RCS temperature increase above the CTS MODE 6 temperature limit while a reactor vessel head closure bolt is less than fully tensioned. This ITS Note is included only for clarity. It is consistent with the current use of MODES 4 and 5 and does not result in any technical change to the application of the MODES. For consistency with the Notes in ITS Table 1.1-1, the ITS definition of MODE adds, "reactor vessel head closure bolt tensioning" to the list of characteristics that define a MODE. Currently, the CTS definition does not include this clarification. This change is acceptable because the definition of MODE should be consistent with the MODE table in order to avoid confusion. This change is made only for consistency and results in no technical changes to the Technical Specifications. These changes are designated as administrative because they clarify the application of the MODES and no technical changes to the MODE definitions are made. The clarifications are consistent with the current use and application of the MODES. A12 CTS Section 1.0 provides a definition of SHUTDOWN MARGIN (SDM). The ITS Section 1.1 definition of SDM contains two differences from the CTS definition. The CTS definition of SDM does not include a statement requiring an increased allowance for the withdrawn worth of an immovable or untrippable control rod(s). This requirement is contained in CTS 4.1.1.1.1.a and CTS 4.1.1.2.a. The ITS definition of SDM includes this increased allowance by stating, "With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM." This changes the CTS definition of SDM to include the requirement in CTS 4.1.1.1.1.a and CTS 4.1.1.2.a for an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s). This change is acceptable because it is consistent with the existing SDM requirements in CTS 3.1.1.1 and 3.1.1.2. The CTS definition is clarified to include a description of the reactor fuel and moderator temperature conditions (i.e., nominal zero power level) at which the SDM is calculated when in MODE 1 or 2. This change is acceptable because including this information is not a technical change. SDM calculations are currently performed for nominal zero power conditions. These changes are designated as administrative because they do not represent a technical change to the Technical Specifications. DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 8 of 11 A13 The CTS Section 1.0 definition of STAGGERED TEST BASIS states, "A STAGGERED TEST BASIS shall consist of: a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval." The ITS Section 1.1 definition states, "A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function." This changes the CTS to specify the frequency of a Surveillance on one system, subsystem, train, or other designated component in the Frequency column of the ITS instead of specifying the frequency in which all systems, subsystems, trains, or other designated components must be tested. This change is acceptable because the testing frequency of components on a STAGGERED TEST BASIS is not changed. Unlike the CTS definition, the ITS definition allows the Surveillance interval for one subsystem to be specified in the Frequency column of the applicable Surveillance Requirements, independent of the number of subsystems. As an example, consider a three-channel system tested on a STAGGERED TEST BASIS. The CTS would specify testing every three months on a STAGGERED TEST BASIS, which results in one channel being tested each month (three equal subintervals). Under the ITS definition, the Surveillance Frequency would be monthly on a STAGGERED TEST BASIS and, one channel would be tested each month. In both the CTS and ITS definitions, all channels are tested every three months. Each test under the CTS definition would be performed at the beginning of the subinterval. Under the ITS definition, each Surveillance Frequency starts at the beginning of the CTS definition subinterval. Thus, there are no net changes in the testing interval. This change represents an editorial preference in the ITS. This change is designated as administrative as no technical changes are made to the Technical Specifications. A14 CTS Table 1.1, OPERATIONAL MODES, is revised. The corresponding table in ITS Section 1.1 is Table 1.1-1, MODES. The changes to the CTS are: The CTS Table 1.1 minimum average reactor coolant temperature for MODES 1 and 2 is changed from 350°F to "NA" (not applicable) in ITS Table 1.1-1. This change is acceptable because ITS LCO 3.4.2, RCS Minimum Temperature for Criticality, provides the minimum reactor coolant temperature limits for MODES 1 and 2. Therefore, the 350°F minimum temperature does not provide any useful information in ITS Table 1.1-1, and is deleted from the CTS. The CTS Table 1.1 MODE 6 upper limit on average reactor coolant temperature (< 140°F) is removed. In ITS Table 1.1-1, the MODE 6 average reactor coolant temperature limit is specified as "NA" (not applicable). DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 9 of 11 This change is acceptable because it eliminates a conflict in the CTS MODE Table. If the average coolant temperature exceeds the upper limit with the reactor vessel head closure bolts less than fully tensioned, the CTS Table could be misinterpreted as no MODE being applicable. This is not the intent of the CTS or ITS MODE 6 definitions. By removing the temperature reference, this ambiguity is eliminated. The CTS Table 1.1 % RATED THERMAL POWER limit of 0% for MODES 3, 4, 5, and 6 is changed in ITS Table 1.1-1 to "NA" (not applicable). This change is acceptable because the reactivity and plant equipment limitations in MODES 3, 4, 5, and 6 do not allow power operation. Therefore, it is not necessary to have these restrictions in the MODE Table. CTS Table 1.1 contains the unit designators of percent (%) and degrees Fahrenheit (°F) next to the values. This is changed in ITS Table 1.1-1 by removing the designator from the individual value(s). This change is acceptable because the designators are contained in the labels associated with the columns. Therefore, it is not necessary to have these designators in the MODE Table. These changes are designated as administrative because they result in no technical changes to the Technical Specifications. A15 ITS Sections 1.2, 1.3, and 1.4 contain information that is not in the CTS. This change to the CTS adds explanatory information on ITS usage that is not applicable to the CTS. The added sections are: Section 1.2 - Logical Connectors Section 1.2 provides specific examples of the logical connectors "AND" and "OR" and the numbering sequence associated with their use. Section 1.3 - Completion Times Section 1.3 provides guidance on the proper use and interpretation of Completion Times. The section also provides specific examples that aid in the use and understanding of Completion Times Section 1.4 - Frequency Section 1.4 provides guidance on the proper use and interpretation of Surveillance Frequencies. The section also provides specific examples that aid in the use and understanding of Surveillance Frequency. This change is acceptable because it aids in the understanding and use of the format and presentation style of the ITS. The addition of these sections does not add or delete technical requirements, and will be discussed specifically in those DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 10 of 11 Technical Specifications where application of the added sections results in a change. This change is designated as administrative because it does not result in a technical change to the Technical Specifications.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS Table 1.1, "OPERATIONAL MODES," states that MODE 6 is restricted to reactivity conditions with keff 0.95. ITS Table 1.1-1, "MODES," does not contain this restriction.

This change is acceptable because the core reactivity requirements for MODE 6 are covered in ITS 3.9.1, "Boron Concentration," by requiring the boron concentration in the Reactor Coolant System to be maintained within the limits specified in the COLR. The LCO section of the 3.9.1 Bases states "The boron concentration limit specified in the COLR ensures that a core keff of 0.95 is maintained during fuel handling operations." Moving this detail from the MODE Table to the LCO 3.9.1 Bases eliminates the potential to misinterpret the MODE table and not apply the MODE 6 requirements if the reactor vessel head closure bolts are less than fully tensioned, fuel is in the reactor vessel, and core reactivity exceeds a keff of 0.95. ITS LCO 3.9.1 will ensure that the appropriate reactivity conditions are maintained in MODE 6, so it is not necessary to have this restriction in the MODE Table in order to provide adequate protection of the public health and safety. Once moved to the Bases, any changes to the core reactivity requirement will be controlled by the Technical Specifications Bases Control Program described in Chapter 5 of the ITS. This change is designated a less restrictive removal of detail because it moves information from the Technical Specifications to the Bases. LA02 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program). CTS 1.14 and CTS Table 1.2 present Frequency Notation for the performance of Surveillance Requirements in the CTS. The ITS specify the periodic Frequency as "In accordance with the Frequency Control Program." This changes the CTS by moving the Frequency Notation Table to the Surveillance Frequency Control Program. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure DISCUSSION OF CHANGES ITS 1.0, USE AND APPLICATIONS Sequoyah Unit 1 and 2 Page 11 of 11 that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to ensure the associated Limiting Conditions for Operations are met. This change is designated as a less restrictive removal of detail change because the Surveillance Frequencies are being removed from the Technical Specifications and placed in a license control document. LESS RESTRICTIVE CHANGES L01 The CTS Section 1.0 definition of CHANNEL FUNCTIONAL TEST requires the use of a simulated signal when performing the test. ITS Section 1.1 renames the CTS definition to CHANNEL OPERATIONAL TEST (COT) (discussed in DOC A04) and allows the use of a simulated or actual signal when performing the test. This changes the CTS by allowing the use of unplanned actuations to perform the Surveillance based on the collection of sufficient information to satisfy the surveillance test requirements. This change is acceptable because the channel itself cannot discriminate between an "actual" or "simulated" signal. Therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change is designated as less restrictive because it allows an actual signal to be credited for Surveillance where only a simulated signal was previously allowed. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Definitions 1.1 Westinghouse STS 1.1-1 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 1.0 USE AND APPLICATION

1.1 Definitions

NOTE----------------------------------------------------------- The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices. AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals (AFD) between the [top and bottom halves of a two section excore neutron detector]. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. 1.0 1.1 1.2 1.4 1.5 CTS 2 Definitions 1.1 Westinghouse STS1.1-2Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 1.1 Definitions CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps. CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity"]. - AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > [15] minutes, making up at least 95% of the total noniodine activity in the coolant. 1.61.10 1.11 TSTF-490 TSTF-490 TSTF-490 INSERT 1 Chapter 1.0 Insert Page 1.1-2a INSERT 1 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using

--------------------------- Reviewer's Note -------------------------------  The thyroid dose conversion factors to be listed are those assumed in the steam generator tube rupture analysis and, if limiting, the steam line break analysis and must be those factors used to calculate the limit in LCO 3.4.16, "RCS Specific Activity."  The first set of thyroid dose conversion factors shall be used for plants licensed to 10 CFR 100.11. The following Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) conversion factors should be used for plants licensed to 10 CFR 50.67.    ---------------------------------------------------------------------------------  
[thyroid dose conversion factors from:  a. Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or  b. Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or  c. ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or  d. Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

OR Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.] TSTF-490 Chapter 1.0 Insert Page 1.1-2b INSERT 1 (continued) DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides [Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138] actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using [effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations" or similar source]. TSTF-490 Definitions 1.1 Westinghouse STS 1.1-3 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 1.1 Definitions

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE); b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE, and

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. 1.13 1.16 1.16 1.37 1.22 ;;; 444 Definitions 1.1 Westinghouse STS 1.1-4 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 1.1 Definitions

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter [14, Initial Test Program] of the FSAR,

b. Authorized under the provisions of 10 CFR 50.59, or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates and the low temperature overpressure protection arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.4. s and Operations, 21.20 1.19 1.21 1.23 ;; 44DOC A02 .0U Definitions 1.1 Westinghouse STS 1.1-5 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 1.1 Definitions

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of [2893] MWt. REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the [nominal zero power design level].

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. 3455 221.26 1.27 1.28 1.31 ;4DOC A02 6 Definitions 1.1 Westinghouse STS 1.1-6 Rev. 4.0 CTS 1SEQUOYAH UNIT 1 Amendment XXX 1.1 Definitions

[ STAGGERED TEST BASIS  A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. ]

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating OPERATIONAL TEST device and verifying the OPERABILITY of all devices in the (TADOT) channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. 21.36 1.35 2DOC A02 Definitions 1.1 Westinghouse STS 1.1-7 Rev. 4.0 CTS 1SEQUOYAH UNIT 1 Amendment XXX Table 1.1-1 (page 1 of 1) MODES MODE TITLE REACTIVITY CONDITION (keff) % RATED THERMAL POWER(a) AVERAGE REACTOR COOLANT TEMPERATURE (°F) 1 Power Operation 0.99 > 5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA [350] 4 Hot Shutdown(b) < 0.99 NA [350] > Tavg > [200] 5 Cold Shutdown(b) < 0.99 NA [200] 6 Refueling(c) NA NA NA

(a) Excluding decay heat. (b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned. 2Table 1.1 22 Logical Connectors 1.2 Westinghouse STS 1.2-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 11.0 USE AND APPLICATION

1.2 Logical Connectors

PURPOSE The purpose of this section is to explain the meaning of logical connectors. Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings. BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors. When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency. EXAMPLES The following examples illustrate the use of logical connectors. Logical Connectors 1.2 Westinghouse STS 1.2-2 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 11.2 Logical Connectors

EXAMPLES (continued)

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met.

A.1 Verify . . .

AND A.2 Restore . . . In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed. Logical Connectors 1.2 Westinghouse STS 1.2-3 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 11.2 Logical Connectors

EXAMPLES (continued)

EXAMPLE 1.2-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met.

A.1 Trip . . .

OR A.2.1 Verify . . . AND A.2.2.1 Reduce . . . OR A.2.2.2 Perform . . .

OR A.3 Align . . .

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed. Completion Times 1.3 Westinghouse STS 1.3-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.0 USE AND APPLICATION

1.3 Completion Times

PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s). DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability. If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition. Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability: a. Must exist concurrent with the first inoperability and Completion Times 1.3 Westinghouse STS 1.3-2 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

DESCRIPTION (continued)

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications. The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 5. 6 hours

36 hours Completion Times 1.3 Westinghouse STS 1.3-3 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. The Required Actions of Condition B are to be in MODE 3 within 6 hours AND in MODE 5 within 36 hours. A total of 6 hours is allowed for reaching MODE 3 and a total of 36 hours (not 42 hours) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours, the time allowed for reaching MODE 5 is the next 33 hours because the total time allowed for reaching MODE 5 is 36 hours. If Condition B is entered while in MODE 3, the time allowed for reaching MODE 5 is the next 36 hours. EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump inoperable. A.1 Restore pump to OPERABLE status. 7 days B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 5. 6 hours

36 hours When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated. Completion Times 1.3 Westinghouse STS 1.3-4 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered. While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A. While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired. On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days. Completion Times 1.3 Westinghouse STS 1.3-5 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

EXAMPLE 1.3-3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Function X train inoperable. A.1 Restore Function X train to OPERABLE status. 7 days B. One Function Y train inoperable. B.1 Restore Function Y train to OPERABLE status. 72 hours

C. One Function X train inoperable. AND One Function Y train inoperable. C.1 Restore Function X train to OPERABLE status. OR C.2 Restore Function Y train to OPERABLE status. 72 hours

72 hours When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered). Completion Times 1.3 Westinghouse STS 1.3-6 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A). It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended. EXAMPLE 1.3-4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more valves inoperable. A.1 Restore valve(s) to OPERABLE status. 4 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 4.

6 hours

12 hours A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times. Completion Times 1.3 Westinghouse STS 1.3-7 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours. If the Completion Time of 4 hours (including the extension) expires while one or more valves are still inoperable, Condition B is entered. EXAMPLE 1.3-5 ACTIONS

--------------------------------------------- NOTE -------------------------------------------  Separate Condition entry is allowed for each inoperable valve.  --------------------------------------------------------------------------------------------------   CONDITION  REQUIRED ACTION  COMPLETION TIME   A. One or more valves inoperable. A.1 Restore valve to OPERABLE status.

4 hours

B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 4. 6 hours

12 hours The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table. (s)5 Completion Times 1.3 Westinghouse STS 1.3-8 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve. If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel inoperable. A.1 Perform SR 3.x.x.x.

OR A.2 Reduce THERMAL POWER to 50% RTP. Once per 8 hours

8 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

6 hours Completion Times 1.3 Westinghouse STS 1.3-9 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed, and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered. If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A. EXAMPLE 1.3-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable. A.1 Verify affected subsystem isolated.

AND A.2 Restore subsystem to OPERABLE status. 1 hour

AND Once per 8 hours thereafter

72 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 5.

6 hours

36 hours Completion Times 1.3 Westinghouse STS 1.3-10 Rev. 4.0 1SEQUOYAH UNIT 1 Amendment XXX 1.3 Completion Times EXAMPLES (continued)

Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.1. If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour or any subsequent 8 hour interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired. IMMEDIATE When "Immediately" is used as a Completion Time, The Required Action COMPLETION TIME should be pursued without delay and in a controlled manner. 5 Frequency 1.4 Westinghouse STS 1.4-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.0 USE AND APPLICATION

1.4 Frequency

PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements. DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0.2, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the Surveillance column that modify performance requirements. Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance or both. Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction. The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria. Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied: 3 Frequency 1.4 Westinghouse STS 1.4-2 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

DESCRIPTION (continued)

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered, or b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed, or c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed. Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations. EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

Frequency 1.4 Westinghouse STS 1.4-3 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable. If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable. Frequency 1.4 Westinghouse STS 1.4-4 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours after 25% RTP AND 24 hours thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP. Frequency 1.4 Westinghouse STS 1.4-5 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Not required to be performed until 12 hours after 25% RTP. ----------------------------------------------------------------

Perform channel adjustment.

7 days The interval continues, whether or not the unit operation is < 25% RTP between performances. As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches 25% RTP to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours (plus the extension allowed by SR 3.0.2) with power 25% RTP. Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply. Frequency 1.4 Westinghouse STS 1.4-6 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Only required to be met in MODE 1. ----------------------------------------------------------------

Verify leakage rates are within limits.

24 hours

Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. Frequency 1.4 Westinghouse STS 1.4-7 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Only required to be performed in MODE 1. ----------------------------------------------------------------

Perform complete cycle of the valve.

7 days

The interval continues, whether or not the unit operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances. As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1. Once the unit reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply. Frequency 1.4 Westinghouse STS 1.4-8 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Not required to be met in MODE 3. ----------------------------------------------------------------

Verify parameter is within limits.

24 hours

Example 1.4-[6] specifies that the requirements of this Surveillance do not have to be met while the unit is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), and the unit was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. 25was Definitions 1.1 Westinghouse STS 1.1-1 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 1.0 USE AND APPLICATION

1.1 Definitions

NOTE----------------------------------------------------------- The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices. AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals (AFD) between the [top and bottom halves of a two section excore neutron detector]. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. 1.0 1.1 1.2 1.4 1.5 CTS 2 Definitions 1.1 Westinghouse STS1.1-2Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 1.1 Definitions CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps. CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity"]. - AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > [15] minutes, making up at least 95% of the total noniodine activity in the coolant. 1.61.10 1.11 TSTF-490 TSTF-490 TSTF-490 INSERT 1 Chapter 1.0 Insert Page 1.1-2a INSERT 1 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using

--------------------------- Reviewer's Note -------------------------------  The thyroid dose conversion factors to be listed are those assumed in the steam generator tube rupture analysis and, if limiting, the steam line break analysis and must be those factors used to calculate the limit in LCO 3.4.16, "RCS Specific Activity."  The first set of thyroid dose conversion factors shall be used for plants licensed to 10 CFR 100.11. The following Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) conversion factors should be used for plants licensed to 10 CFR 50.67.    ---------------------------------------------------------------------------------  
[thyroid dose conversion factors from:  a. Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or  b. Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or  c. ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or  d. Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

OR Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.] TSTF-490 Chapter 1.0 Insert Page 1.1-2b INSERT 1 (continued) DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides [Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138] actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using [effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations" or similar source]. TSTF-490 Definitions 1.1 Westinghouse STS 1.1-3Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE1.LEAKAGE, such as that from pump seals or valvepacking (except reactor coolant pump (RCP) sealwater injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,2.LEAKAGE into the containment atmosphere fromsources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or 3.Reactor Coolant System (RCS) LEAKAGEthrough a steam generator to the Secondary System (primary to secondary LEAKAGE); b. Unidentified LEAKAGEAll LEAKAGE (except RCP seal water injection orleakoff) that is not identified LEAKAGE, andc.Pressure Boundary LEAKAGELEAKAGE (except primary to secondary LEAKAGE)through a nonisolable fault in an RCS component body,pipe wall, or vessel wall.1.13 1.16 1.16 1.37 1.22 ;;; 444 Definitions 1.1 Westinghouse STS 1.1-4 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 1.1 Definitions

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter [14, Initial Test Program] of the FSAR,

b. Authorized under the provisions of 10 CFR 50.59, or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates and the low temperature overpressure protection arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.4. s and Operations, 21.20 1.19 1.21 1.23 ;; 44DOC A02 .0U Definitions 1.1 Westinghouse STS 1.1-5 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 1.1 Definitions

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of [2893] MWt. REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the [nominal zero power design level].

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. 3455 221.26 1.27 1.28 1.31 ;4DOC A02 6 Definitions 1.1 Westinghouse STS 1.1-6 Rev. 4.0 CTS 1SEQUOYAH UNIT 2 Amendment XXX 1.1 Definitions

[ STAGGERED TEST BASIS  A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. ]

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating OPERATIONAL TEST device and verifying the OPERABILITY of all devices in the (TADOT) channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. 21.36 1.35 2DOC A02 Definitions 1.1 Westinghouse STS 1.1-7 Rev. 4.0 CTS 1SEQUOYAH UNIT 2 Amendment XXX Table 1.1-1 (page 1 of 1) MODES MODE TITLE REACTIVITY CONDITION (keff) % RATED THERMAL POWER(a) AVERAGE REACTOR COOLANT TEMPERATURE (°F) 1 Power Operation 0.99 > 5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA [350] 4 Hot Shutdown(b) < 0.99 NA [350] > Tavg > [200] 5 Cold Shutdown(b) < 0.99 NA [200] 6 Refueling(c) NA NA NA

(a) Excluding decay heat. (b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned. 2Table 1.1 22 Logical Connectors 1.2 Westinghouse STS 1.2-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 11.0 USE AND APPLICATION

1.2 Logical Connectors

PURPOSE The purpose of this section is to explain the meaning of logical connectors. Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings. BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors. When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency. EXAMPLES The following examples illustrate the use of logical connectors. Logical Connectors 1.2 Westinghouse STS 1.2-2 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 11.2 Logical Connectors

EXAMPLES (continued)

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met.

A.1 Verify . . .

AND A.2 Restore . . . In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed. Logical Connectors 1.2 Westinghouse STS 1.2-3 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 11.2 Logical Connectors

EXAMPLES (continued)

EXAMPLE 1.2-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met.

A.1 Trip . . .

OR A.2.1 Verify . . . AND A.2.2.1 Reduce . . . OR A.2.2.2 Perform . . .

OR A.3 Align . . .

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed. Completion Times 1.3 Westinghouse STS 1.3-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.0 USE AND APPLICATION

1.3 Completion Times

PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s). DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability. If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition. Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability: a. Must exist concurrent with the first inoperability and Completion Times 1.3 Westinghouse STS 1.3-2 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

DESCRIPTION (continued)

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications. The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 5. 6 hours

36 hours Completion Times 1.3 Westinghouse STS 1.3-3 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. The Required Actions of Condition B are to be in MODE 3 within 6 hours AND in MODE 5 within 36 hours. A total of 6 hours is allowed for reaching MODE 3 and a total of 36 hours (not 42 hours) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours, the time allowed for reaching MODE 5 is the next 33 hours because the total time allowed for reaching MODE 5 is 36 hours. If Condition B is entered while in MODE 3, the time allowed for reaching MODE 5 is the next 36 hours. EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump inoperable. A.1 Restore pump to OPERABLE status. 7 days B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 5. 6 hours

36 hours When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated. Completion Times 1.3 Westinghouse STS 1.3-4 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered. While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A. While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired. On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days. Completion Times 1.3 Westinghouse STS 1.3-5 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

EXAMPLE 1.3-3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Function X train inoperable. A.1 Restore Function X train to OPERABLE status. 7 days B. One Function Y train inoperable. B.1 Restore Function Y train to OPERABLE status. 72 hours

C. One Function X train inoperable. AND One Function Y train inoperable. C.1 Restore Function X train to OPERABLE status. OR C.2 Restore Function Y train to OPERABLE status. 72 hours

72 hours When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered). Completion Times 1.3 Westinghouse STS 1.3-6 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A). It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended. EXAMPLE 1.3-4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more valves inoperable. A.1 Restore valve(s) to OPERABLE status. 4 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 4.

6 hours

12 hours A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times. Completion Times 1.3 Westinghouse STS 1.3-7 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours. If the Completion Time of 4 hours (including the extension) expires while one or more valves are still inoperable, Condition B is entered. EXAMPLE 1.3-5 ACTIONS

--------------------------------------------- NOTE -------------------------------------------  Separate Condition entry is allowed for each inoperable valve.  --------------------------------------------------------------------------------------------------   CONDITION  REQUIRED ACTION  COMPLETION TIME   A. One or more valves inoperable. A.1 Restore valve to OPERABLE status.

4 hours

B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 4. 6 hours

12 hours The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table. (s)5 Completion Times 1.3 Westinghouse STS 1.3-8 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve. If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel inoperable. A.1 Perform SR 3.x.x.x.

OR A.2 Reduce THERMAL POWER to 50% RTP. Once per 8 hours

8 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

6 hours Completion Times 1.3 Westinghouse STS 1.3-9 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.3 Completion Times

EXAMPLES (continued)

Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed, and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered. If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A. EXAMPLE 1.3-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable. A.1 Verify affected subsystem isolated.

AND A.2 Restore subsystem to OPERABLE status. 1 hour

AND Once per 8 hours thereafter

72 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 5.

6 hours

36 hours Completion Times 1.3 Westinghouse STS 1.3-10 Rev. 4.0 1SEQUOYAH UNIT 2 Amendment XXX 1.3 Completion Times EXAMPLES (continued)

Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.1. If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour or any subsequent 8 hour interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired. IMMEDIATE When "Immediately" is used as a Completion Time, The Required Action COMPLETION TIME should be pursued without delay and in a controlled manner. 5 Frequency 1.4 Westinghouse STS 1.4-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.0 USE AND APPLICATION

1.4 Frequency

PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements. DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0.2, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the Surveillance column that modify performance requirements. Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance or both. Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction. The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria. Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied: 3 Frequency 1.4 Westinghouse STS 1.4-2 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

DESCRIPTION (continued)

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered, or b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed, or c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed. Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations. EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

Frequency 1.4 Westinghouse STS 1.4-3 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable. If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable. Frequency 1.4 Westinghouse STS 1.4-4 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours after 25% RTP AND 24 hours thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP. Frequency 1.4 Westinghouse STS 1.4-5 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Not required to be performed until 12 hours after 25% RTP. ----------------------------------------------------------------

Perform channel adjustment.

7 days The interval continues, whether or not the unit operation is < 25% RTP between performances. As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches 25% RTP to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours (plus the extension allowed by SR 3.0.2) with power 25% RTP. Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply. Frequency 1.4 Westinghouse STS 1.4-6 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Only required to be met in MODE 1. ----------------------------------------------------------------

Verify leakage rates are within limits.

24 hours

Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. Frequency 1.4 Westinghouse STS 1.4-7 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Only required to be performed in MODE 1. ----------------------------------------------------------------

Perform complete cycle of the valve.

7 days

The interval continues, whether or not the unit operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances. As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1. Once the unit reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply. Frequency 1.4 Westinghouse STS 1.4-8 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX11.4 Frequency

EXAMPLES (continued)

EXAMPLE 1.4-6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE--------------------------- Not required to be met in MODE 3. ----------------------------------------------------------------

Verify parameter is within limits.

24 hours

Example 1.4-[6] specifies that the requirements of this Surveillance do not have to be met while the unit is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), and the unit was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. 25was JUSTIFICATION FOR DEVIATIONS ITS 1.0, USE AND APPLICATION Sequoyah Unit 1 and 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plantspecific nomenclature, number, reference, system description, analysis, or licensing basisdescription.2.The ISTS contains bracketed information and/or values that are generic to all Westinghousevintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the currentlicensing basis.3.Typographical error is corrected. The proper section for Surveillance Requirement (SR)Applicability is Section 3.0.4.These punctuation corrections have been made consistent with the Writers Guide for theImproved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.5.Typographical error is corrected.6.The ISTS definition of Shutdown Margin states in part, "However, with all RCCAs verifiedfully inserted by two independent means, it is not necessary to account for a stuck RCCA inthe SDM calculation." The CTS definition of Shutdown Margin does not contain this allowance, therefore the ITS does not include this allowance. This is acceptable since theinformation is changed to reflect the current licensing basis. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 1.0, USE AND APPLICATION Sequoyah Unit 1 and 2 Page 1 of 2 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L01 The Tennessee Valley Authority (TVA) is converting Sequoyah to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, Rev. 4, "Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the determination of no significant hazards considerations for conversion to NUREG-1431. The CTS Section 1.0 definition of CHANNEL FUNCTIONAL TEST requires the use of a simulated signal when performing the test. ITS Section 1.1 renames the CTS definition to CHANNEL OPERATIONAL TEST (COT) and allows the use of a simulated or actual signal when performing the test. This changes the CTS by allowing the use of unplanned actuations to perform the Surveillance based on the collection of sufficient information to satisfy the surveillance test requirements.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed change adds an allowance that an actual as well as a simulated signal can be credited during the COT. This change allows taking credit for unplanned actuations if sufficient information is collected to satisfy the surveillance test requirements. This change is acceptable because the channel itself cannot discriminate between an "actual" or "simulated" signal, and the proposed requirement does not change the technical content or validity of the test. This change will not affect the probability of an accident. The source of the signal sent to components during a Surveillance is not assumed to be an initiator of any analyzed event. The consequence of an accident is not affected by this change. The results of the testing, and, therefore, the likelihood of discovering an inoperable component, are unaffected. As a result, the assurance that equipment will be available to mitigate the consequences of an accident is unaffected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 1.0, USE AND APPLICATION Sequoyah Unit 1 and 2 Page 2 of 2 The proposed change adds an allowance that an actual as well as a simulated signal can be credited during the COT. This change will not physically alter the plant (no new or different type of equipment will be installed). The change does not require any new or revised operator actions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change adds an allowance that an actual as well as a simulated signal can be credited during the COT. The margin of safety is not affected by this change. This change allows taking credit for unplanned actuations if sufficient information is collected to satisfy the surveillance test requirements. This change is acceptable because the channel itself cannot discriminate between an "actual" or "simulated" signal. As a result, the proposed requirement does not change the technical content or validity of the test.

Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, TVA concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified. _ ENCLOSURE 2 VOLUME 4 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS Revision 0 LIST OF ATTACHMENTS 1. ITS Chapter 2.0, Safety Limits ATTACHMENT 1 ITS Chapter 2.0, SAFETY LIMITS (SLs) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS Chapter 2.0 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded:. 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. 2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained 4901°F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications. APPLICABILITY: MODES 1 and 2. ACTION: If SL 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. September 26, 2012 SEQUOYAH - UNIT 1 2-1 Amendment No. 41, 331 2.1 2.1.1 Applicability 2.2.1 2.2.2.1 2.2.2.2 in the COLR LA01Page 1 of 6 2.1.2 2.2.2.2 2.2.2.1 Applicability A01ITS ITS Chapter 2.0 Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation September 26, 2012 SEQUOYAH - UNIT 1 2-2 Amendment No. 19, 331 LA01Page 2 of 6 A01ITS ITS Chapter 2.0 This page deleted. September 3, 1985 SEQUOYAH - UNIT 1 2-3 Amendment No. 41 Page 3 of 6 A01ITS Chapter 2.0 ITS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. 2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained 4901°F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications. APPLICABILITY: MODES 1 and 2. ACTION: If SL 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. SEQUOYAH - UNIT 2 2-1 September 26, 2012 Amendment No. 33, 324 2.1 2.1.1 Applicability 2.2.1 2.1.2 2.2.2.1 2.2.2.1 2.2.2.2 2.2.2.2 in the COLR LA01Page 4 of 6 Applicability A01ITS Chapter 2.0 ITS Figure 2.1-1 Reactor Core Safety Limit-Four Loops in Operation FRACTION OF RATED THERMAL POWER September 26, 2012 SEQUOYAH - UNIT 2 2-2 Amendment No. 21, 324 LA01Page 5 of 6 A01ITS Chapter 2.0 ITS This page deleted. September 3, 1985 SEQUOYAH - UNIT 2 2-3 Amendment No. 33 Page 6 of 6 DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SLs) Sequoyah Unit 1 and Unit 2 Page 1 of 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 6 - Removal of Cycle - Specific Limits from the Technical Specifications to the Core Operating Limits Report) CTS 2.1.1 requires the combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) not to exceed the limits shown in Figure 2.1-1. ITS 2.1.1 states the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR. This changes the CTS by moving limits that must be confirmed on a cycle specific bases to the COLR. The Reactor Core safety limits are retained in Technical Specification Chapter 2.0. The removal of these cycle specific parameter limits from the Technical Specifications to the COLR and the retention of the limiting Safety Limits in the Technical Specifications is acceptable because the cycle specific limits are developed or utilized under NRC-approved methodologies that ensure the Safety Limits are met. The NRC documented in Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," that this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the Safety Limits. NRC-approved Topical Report WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report," determined that the specific values for these parameters may be relocated to the COLR provided the SLs continue to appear in the Technical Specifications. The methodologies used to develop the parameters in the COLR were approved by the NRC in accordance with Generic Letter 88-16. Additionally, this change is acceptable because the removed information will be adequately controlled in the COLR DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SLs) Sequoyah Unit 1 and Unit 2 Page 2 of 2 under the requirements provided in ITS 5.6.3, "Core Operating Limits Report." ITS 5.6.3 ensures that the applicable limits of the safety analysis are met (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such as SDM, transient analysis limits, and accident analysis limits). This change is designated as a less restrictive removal of detail change because information relating to cycle specific parameter limits is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) SLs 2.0 Westinghouse STS 2.0-1Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained [1.17 for the WRB-1/WRB-2 DNB correlations]. 2.1.1.2 The peak fuel centerline temperature shall be maintained < [5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup]. 2.1.2 Reactor Coolant System Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained [2735] psig. 2.2 SAFETY LIMIT VIOLATIONS 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.1.1, 2.1.1 Applicability 2.1 2.1.2, 2.1.2 Applicability 2.1.1 ACTION 2.1.2 ACTION 2.1.2 ACTION 2.1.2 ACTION 111maximum local fuel pin 22.1.1.1 2.1.1.2 INSERT 2INSERT 1 2.1.1 Insert Page 2.0-1 CTS INSERT 1 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. INSERT 2 4901°F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications 11 , Volume 4, Rev 1, Page 15a of 38 Enclosure 2, Volume 4, Rev 1, Page 15a of 38 2.1.1INSERT3Figure2.11ReactorCoreSafetyLimitFourLoopsinOperationFRACTIONOFRATEDTHERMALPOWER SLs 2.0 Westinghouse STS2.0-1Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained [1.17 for the WRB-1/WRB-2 DNB correlations]. 2.1.1.2 The peak fuel centerline temperature shall be maintained < [5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup]. 2.1.2 Reactor Coolant System Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained [2735] psig. 2.2 SAFETY LIMIT VIOLATIONS 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.1.1, 2.1.1 Applicability 2.1 2.1.2, 2.1.2 Applicability 2.1.1 ACTION 2.1.2 ACTION 2.1.2 ACTION 2.1.2 ACTION 111maximum local fuel pin 22.1.1.1 2.1.1.2 INSERT 2INSERT 1 2.1.1 Insert Page 2.0-1 CTS INSERT 1 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. INSERT 2 4901°F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications 11 , Volume 4, Rev 1, Page 17a of 38 Enclosure 2, Volume 4, Rev 1, Page 17a of 38 2.1.1INSERT3Figure2.11ReactorCoreSafetyLimitFourLoopsinOperationFRACTIONOFRATEDTHERMALPOWER JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0, SAFETY LIMITS (SLs) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.The ISTS contains bracketed information and/or values that are generic toWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is inserted to reflect the current licensing basis.2.Changes are made (additions, deletions, and/or changes) to the ISTS that reflect theplant specific nomenclature, number, reference, system description, analysis, orlicensing basis description. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Reactor Core SLs B 2.1.1 Westinghouse STS B 2.1.1-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core

BASES BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature. The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant. The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs. INSERT 1 INSERT 2INSERT 3111corresponding significant B 2.0 Insert Pages B 2.1.1-1a INSERT 1 (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt(CFM)), either of which could result in INSERT 2 from the outer surface of the cladding to the reactor coolant water INSERT 3 DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty. Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the value acceptance criteria in the safety analysis. 111 B 2.0 Insert Pages B 2.1.1-1b INSERT 3 (cont) The curves provided in the COLR show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These lines are bounding for all fuel types. The curves provided in the COLR are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety Limit System Settings (Reactor Trip System trip limits). The plant trip set points are verified to be less than the limits defined by the safety limit lines provided in the COLR converted from power to delta-temperature and adjusted for uncertainty. The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (I) is within the limits of the f1 (Delta I) function of the Overtemperature Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f1(l) trip reset function, the Overtemperature Delta Temperature trip set point is reduced by the values in the COLR to provide protection required by the core safety limits. Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (I) is within the limits of the f2(I) function of the Overpower-Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(I) trip reset function, the Overpower-Delta Temperature trip set point is reduced by the values specified in the COLR to provide protection required by the core safety limits. 1 Reactor Core SLs B2.1.1Westinghouse STS B 2.1.1-2Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the following fuel design criteria: a.There must be at least 95% probability at a 95% confidence level (the95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB andb.The hot fuel pellet in the core must not experience centerline fuelmelting.The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, I, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities. Automatic enforcement of these reactor core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves. The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded. SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation. The reactor core SLs are established to preclude violation of the following fuel design criteria: a.There must be at least a 95% probability at a 95% confidence level(the 95/95 DNB criterion) that the hot fuel rod in the core does notexperience DNB andb.There must be at least a 95% probability at a 95% confidence levelthat the hot fuel pellet in the core does not experience centerline fuel melting.U11 Reactor Core SLs B2.1.1Westinghouse STSB 2.1.1-3Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES SAFETY LIMITS (continued) The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower T reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and I that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs. APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER. SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS SLs. If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable. The allowed Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage. REFERENCES 1.10 CFR 50, Appendix A, GDC 10.2. FSAR, Section [7.2].U 12 RCS Pressure SL B2.1.2Westinghouse STS B 2.1.2-1Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX B 2.0 SAFETY LIMITS (SLs) B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor pressure coolant boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding. The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the reactor high pressure trip have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded. The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of coolant2485 psig 31 RCS Pressure SL B2.1.2Westinghouse STS B 2.1.2-2Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES APPLICABLE SAFETY ANALYSES (continued) external load without a direct reactor trip. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained. The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs. The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices. More specifically, no credit is taken for operation of any of the following: a.Pressurizer power operated relief valves (PORVs),b.Steam line relief valve,c.Steam Dump System,d.Reactor Control System,e.Pressurizer Level Control System, orf.Pressurizer spray valve.SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under [USAS, Section B31.1 (Ref. 6)] is 120% of design pressure. The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig. 2b c d e 44444; ; 5 RCS Pressure SL B2.1.2Westinghouse STS B 2.1.2-3Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized. SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 VIOLATIONS or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour. Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). The allowable Completion Time of 1 hour recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized. If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress. REFERENCES 1.10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.2.ASME, Boiler and Pressure Vessel Code, Section III,Article NB-7000.3.ASME, Boiler and Pressure Vessel Code, Section XI,Article IWX-5000.4.10 CFR 100.5. FSAR, Section[7.2].6.USAS B31.1, Standard Code for Pressure Piping, American Societyof Mechanical Engineers, 1967.2, 1971U 11 Reactor Core SLs B2.1.1Westinghouse STSB 2.1.1-1Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 B 2.0 SAFETY LIMITS (SLs) B 2.1.1 Reactor Core BASES BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature. The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant. The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs. INSERT 1 INSERT 2INSERT 3111corresponding significant B 2.0 Insert Pages B 2.1.1-1a INSERT 1 (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt(CFM)), either of which could result in INSERT 2 from the outer surface of the cladding to the reactor coolant water INSERT 3 DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty. Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the value acceptance criteria in the safety analysis. 111 B 2.0 Insert Pages B 2.1.1-1b INSERT 3 (cont) The curves provided in the COLR show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These lines are bounding for all fuel types. The curves provided in the COLR are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety Limit System Settings (Reactor Trip System trip limits). The plant trip set points are verified to be less than the limits defined by the safety limit lines provided in the COLR converted from power to delta-temperature and adjusted for uncertainty. The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (I) is within the limits of the f1 (Delta I) function of the Overtemperature Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f1(l) trip reset function, the Overtemperature Delta Temperature trip set point is reduced by the values in the COLR to provide protection required by the core safety limits. Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (I) is within the limits of the f2(I) function of the Overpower-Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(I) trip reset function, the Overpower-Delta Temperature trip set point is reduced by the values specified in the COLR to provide protection required by the core safety limits. 1 Reactor Core SLs B2.1.1Westinghouse STS B 2.1.1-2Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the following fuel design criteria: a.There must be at least 95% probability at a 95% confidence level (the95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB andb.The hot fuel pellet in the core must not experience centerline fuelmelting.The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, I, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities. Automatic enforcement of these reactor core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves. The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded. SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation. The reactor core SLs are established to preclude violation of the following fuel design criteria: a.There must be at least a 95% probability at a 95% confidence level(the 95/95 DNB criterion) that the hot fuel rod in the core does notexperience DNB andb.There must be at least a 95% probability at a 95% confidence levelthat the hot fuel pellet in the core does not experience centerline fuel melting.U11 Reactor Core SLs B 2.1.1 Westinghouse STS B 2.1.1-3 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

SAFETY LIMITS (continued) The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower T reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and I that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs. APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER. SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS SLs. If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable. The allowed Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. 2. FSAR, Section [7.2]. U 12 RCS Pressure SL B2.1.2Westinghouse STS B 2.1.2-1Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX B 2.0 SAFETY LIMITS (SLs) B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor pressure coolant boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding. The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the reactor high pressure trip have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded. The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of coolant2485 psig 31 RCS Pressure SL B 2.1.2 Westinghouse STS B 2.1.2-2 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

APPLICABLE SAFETY ANALYSES (continued) external load without a direct reactor trip. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained. The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs. The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices. More specifically, no credit is taken for operation of any of the following: a. Pressurizer power operated relief valves (PORVs),

b. Steam line relief valve,

c. Steam Dump System, d. Reactor Control System, e. Pressurizer Level Control System, or

f. Pressurizer spray valve. SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under [USAS, Section B31.1 (Ref. 6)] is 120% of design pressure. The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

2b c d e 44444; ; 5 RCS Pressure SL B 2.1.2 Westinghouse STS B 2.1.2-3 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized. SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 VIOLATIONS or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour. Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). The allowable Completion Time of 1 hour recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized. If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress. REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000. 3. ASME, Boiler and Pressure Vessel Code, Section XI, Article IWX-5000.

4. 10 CFR 100.

5. FSAR, Section [7.2].

6. USAS B31.1, Standard Code for Pressure Piping, American Society of Mechanical Engineers, 1967. 2, 1971U 11 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0 BASES, SAFETY LIMITS (SLs) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. Typographical/grammatical error corrected. 4. The steam line relief valves are removed from the list of items that have no credit taken for operation. The steam line safety valves are credited with protecting the Reactor Coolant System and the steam generators against overpressure for all load losses. Additionally, the subsequent items have been renumbered.

5. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS CHAPTER 2.0, SAFETY LIMITS (SLs) Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ENCLOSURE 2 VOLUME 5 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.0 LCO AND SR APPLICABILITY Revision 0 LIST OF ATTACHMENTS 1. ITS Section 3.0, LCO and SR Applicability ATTACHMENT 1 ITS Section 3.0, LCO AND SR APPLICABILITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS ITS Section 3.0 A013/4 LIMITING CONDITIONS FOR OPEATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met and as provided in LCO 3.0.7. 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Conditions for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in: 1. At least HOT STANDBY within the next 6 hours, 2. At least HOT SHUTDOWN within the following 6 hours, and 3. At least COLD SHUTDOWN within the subsequent 24 hours. Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or c. When an allowance is stated in the individual value, parameter, or other Specification. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this Specification. Unless both

October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-1 Amendment No. 202, 301, 312 See ITS 3.8.1 LCOs shall be met3.0.8, and LCO 3.0.9 in the Applicability, (LCO).0 A01LCO 3.0.1 A02INSERT 1A02A03an LCO and SINSERT 2 MODE 3 MODE 4 MODE 5 71337INSERT 3this specification A05A06A01LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 A04LCO 3.0.2, LCO 3.0.7, Page 1 of 14 ; A01 ITS Section 3.0 Insert Page 3/4 0-1 INSERT 1 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. INSERT 2 are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable.

INSERT 3 in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. A02 A03A04A01A06Page 2 of 14 ITS ITS Section 3.0 A01APPLICABILITY LIMITING CONDITION FOR OPERATION (Continued) 3.0.5 (Continued) conditions (1) and (2) are satisfied, within 2 hours action shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in: 1.At least HOT STANDBY within the next 6 hours,2.At least HOT SHUTDOWN within the following 6 hours, and 3.At least COLD SHUTDOWN within the subsequent 24 hours.This Specification is not applicable in MODES 5 or 6. 3.0.6 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.1 and 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. 3.0.7 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and: a.the snubbers not able to perform their associated support function(s) are associated with only onetrain or subsystem of a multiple train or subsystem supported system or are associated with a singletrain or subsystem supported system and are able to perform their associated support function within72 hours; orb.the snubbers not able to perform their associated support function(s) are associated with more thanone train or subsystem of a multiple train or subsystem supported system and are able to performtheir associated support function within 12 hours.At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met. SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the MODES or other specified conditions in the Applicability for individual Limiting Condition for Operation, unless otherwise stated in the individual Surveillance Requirement. Failure to meet a Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limiting Condition for Operation. Failure to perform a Surveillance within the specified surveillance interval shall be failure to meet the Limiting Conditions for Operation except as provided in Specification 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. 4.0.3 If it is discovered that a Surveillance was not performed within its specified surveillance interval (including the allowed extension per Specification 4.0.2), then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-2 Amendment No. 78, 162, 202, 208, 274, 280, 293, 312 See ITS 3.8.1 LCO 3.0.5 A07LCO 3.0.8 A08A09INSERT 4 INSERT 5INSERT 6 L01(SR) APPLICABILITY3.0 SRs LCOsSR LCO Frequency SR 3.0.3 A01SR 3.0.1 SR 3.0.2 SR 3.0.3 INSERT 7A10L02M01LCO Frequency A11A01Page 3 of 14 FrequencyA01 ITS Section 3.0 Insert Page 3/4 0-2a INSERT 4 LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. INSERT 5 LCO 3.0.7 Test Exception LCO 3.1.8, PHYSICS TEST Exceptions - MODE 2" allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. A08A09Page 4 of 14 ITS Section 3.0 Insert Page 3/4 0-2b INSERT 6 LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events. If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s). At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met. INSERT 7 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications.

L01A10M01A10L02Page 5 of 14 ITS ITS Section 3.0 A01SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued) If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows: Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following: a. Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a; b. The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspection activities;

c. Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisions of SR 4.0.2 are not applicable); and d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirement of any TS. Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;

October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-3 Amendment No. 78, 162, 202, 208, 274, 280, 293, 301, 308 SR 3.0.3 A01LCO LCO ConditionCondition SR 3.0.4 See ITS 5.5 Page 6 of 14 ITS ITS Section 3.0 A01APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued) b. Testing Frequencies applicable to the ASME OM Code and applicable Addenda as follows: ASME OM Code and applicable Addenda Required frequencies for terminology for inservice performing inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;

d. The provisions of SR 4.0.3 are applicable to inservice testing and activities; and e. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-4 Amendment No. 78, 162, 202, 208, 274, 280, 293, 308 See ITS 5.5 Page 7 of 14 ITS Section 3.0ITS A013/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met and as provided in LCO 3.0.7. 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Conditions for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in: 1.At least HOT STANDBY within the next 6 hours,2.At least HOT SHUTDOWN within the following 6 hours, and3.At least COLD SHUTDOWN within the subsequent 24 hours.Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a.When the associated ACTIONS to be entered permit continued operation in the MODE or otherspecified condition in the Applicability for an unlimited period of time;b.After performance of a risk assessment addressing inoperable systems and components,consideration of the results, determination of the acceptability of entering the MODE or otherspecified condition in the Applicability, and establishment of risk management actions, ifappropriate; exceptions to this Specification are stated in the individual Specifications, orc.When an allowance is stated in the individual value, parameter, or other Specification.This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this Specification. Unless both October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-1 Amendment No. 192, 290, 301 See ITS 3.8.1 LCOs shall be metin the Applicability, (LCO).0 A01LCO 3.0.1 A02INSERT 1A02A03an LCO and SINSERT 2 MODE 3 MODE 4 MODE 5 71337INSERT 3this specification A05A06A01LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 A043.0.8, and LCO 3.0.9 LCO 3.0.2, LCO 3.0.7, Page 8 of 14 ;A01 ITS Section 3.0 Insert Page 3/4 0-1 INSERT 1 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. INSERT 2 are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. INSERT 3 in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. A02 A03A04A01A06Page 9 of 14 ITS Section 3.0ITS A01APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.5 (Continued) conditions (1) and (2) are satisfied, within 2 hours action shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in: 1. At least HOT STANDBY within the next 6 hours 2. At least HOT SHUTDOWN within the following 6 hours, and

3. At least COLD SHUTDOWN within the subsequent 24 hours. This Specification is not applicable in MODES 5 or 6. 3.0.6 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.1 and 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. 3.0.7 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and: a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours; or b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours. At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met. SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the MODES or other specified conditions in the Applicability for individual Limiting Condition for Operation, unless otherwise stated in the individual Surveillance Requirement. Failure to meet a Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limiting Condition for Operation. Failure to perform a Surveillance within the specified surveillance interval shall be failure to meet the Limiting Conditions for Operation except as provided in Specification 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

4.0.3 If it is discovered that a Surveillance was not performed within its specified surveillance interval (including the allowed extension per Specification 4.0.2), then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-2 Amendment No. 69, 152, 192, 198, 263, 271, 283, 301 See ITS 3.8.1 LCO 3.0.5 A07LCO 3.0.8 A08A09INSERT 4 INSERT 5INSERT 6 L01(SR) APPLICABILITY3.0 SRs LCOsSR LCO Frequency SR 3.0.3 A01SR 3.0.1 SR 3.0.2 SR 3.0.3 INSERT 7A10L02M01LCO A11A01Page 10 of 14FrequencyA01 ITS Section 3.0 Insert Page 3/4 0-2a INSERT 4 LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. INSERT 5 LCO 3.0.7 Test Exception LCO 3.1.8, PHYSICS TEST Exceptions - MODE 2" allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. A08A09Page 11 of 14 ITS Section 3.0 Insert Page 3/4 0-2b INSERT 6 LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events. If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s). At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met. INSERT 7 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. L01A10M01A10L02Page 12 of 14 ITS Section 3.0ITS A01APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued) up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows: Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following: a.Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed inaccordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addendaas required by 10 CFR 50.55a;b.The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspectionactivities;c.Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Positionc.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), aqualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to thecircle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant)of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisionsof SR 4.0.2 are not applicable); andd.Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede therequirement of any TS.Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-3 Amendment No. 69, 152, 198, 263, 271, 283, 290 FrequencyA01SR 3.0.3 A01LCO LCO ConditionCondition SR 3.0.4 See ITS 5.5 Page 13 of 14 ITS Section 3.0ITS A01APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued) a.Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall beperformed in accordance with the ASME Code for Operation and Maintenance of Nuclear PowerPlants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;b.Testing frequencies applicable to the ASME OM Code and applicable Addenda as follows:ASME OMCode and applicable AddendaRequired frequencies for terminology for inserviceperforming inservice testing activitiestesting activities WeeklyAt least once per 7 days MonthlyAt least once per 31 days Quarterly or every 3 monthsAt least once per 92 days Semiannually or every 6 monthsAt least once per 184 days Every 9 monthsAt least once per 276 days Yearly or annuallyAt least once per 366 days Biennially or every 2 yearsAt least once per 731 days c.The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normal andaccelerated frequencies specified as 2 years or less in the Inservice Test Program forperforming inservice testing activities;d.The provisions of SR 4.0.3 are applicable to inservice testing and activities; ande.Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-4 Amendment No. 69, 152, 198, 263, 271, 283, 297 See ITS 5.5 Page 14 of 14 DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 1 of 10 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 Compliance with the Limiting Conditions for Operation (LCO) contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met and as provided in LCO 3.0.7." ITS LCO 3.0.1 states, "LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9." This results in several changes to the CTS.

Certain phrases are revised to be consistent with the equivalent phrase used in the ITS. Specifically, "OPERATIONAL MODES or other conditions specified therein" is changed to "MODES or other specified conditions in the Applicability" to be consistent with the ITS definition of MODE and the terminology used in the ITS.

These changes are acceptable because they result in no change in the intent or application of the Technical Specifications, but merely reflect editorial preferences used in the ITS.

The phrase "Compliance with the LIMITING CONDITIONS FOR OPERATION contained in the succeeding Specifications is required" is replaced with "LCOs shall be met." This change is made consistent with the ITS. This change is acceptable because it is an editorial change that does not change the intent of the requirements.
The phrase "except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met" is moved from CTS 3.0.1 to ITS LCO 3.0.2, which states in part, "Upon discovery or failure to meet an LCO, the Required Actions of the associated Conditions shall be met." The change is acceptable because moving this information within the Technical Specifications results in no change in the intent or application of ACTIONS.

DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 2 of 10 *The phrase "except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, andLCO 3.0.9" is added in ITS LCO 3.0.1. ITS LCO 3.0.2 describes theappropriate actions to be taken when ITS LCO 3.0.1 is not met. ITSLCO 3.0.7 describes Test Exceptions LCOs, which are exceptions to other LCOs. ITS LCO 3.0.8 addresses snubber inoperabilities, which is also an exception to other LCOs. ITS LCO 3.0.9 addresses barrier inoperabilities which is also an exception to other LCOs. Changes resulting from the incorporation of ITS LCO 3.0.9 are discussed in Discussion of Change(DOC) L01.This change is acceptable because adding the exceptions for ITS LCO 3.0.2, LCO 3.0.6, LCO 3.0.7, and LCO 3.0.9 prevent a conflict within the Applicability section. This addition is needed for consistency in the ITS requirements and does not change the intent or application of the Technical Specifications. Furthermore, changing the CTS LCO 3.0.7 to ITS LCO 3.0.8 does not change the intent of the snubber exception. A03 CTS 3.0.2 states, "Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Conditions for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required." ITS LCO 3.0.2 states "Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated." This results in several changes to the CTS. *CTS 3.0.2 is revised to include an exception for ITS LCO 3.0.6. ITS LCO3.0.6 is a new allowance that takes exception to the ITS LCO 3.0.2requirement to take the associated ACTION requirements when a LIMITINGCONDITION FOR OPERATION is not met. This exception is included in ITS LCO 3.0.2 to avoid conflict between the applicability requirements.This change is acceptable because it includes a reference to a new item inthe ITS and results in no change to the CTS. Changes resulting from theincorporation of ITS LCO 3.0.6 are discussed in DOC A07.*The second sentence of CTS 3.0.2 states, "If the LIMITING CONDITIONSFOR OPERATION is restored prior to expiration of the specified timeintervals, completion of the ACTION requirements is not required." The sentence is changed, in ITS LCO 3.0.2, to state "If the LCO is not met or is no longer applicable prior to expiration of the specified Completion Time(s),completion of the Required Action(s) is not required unless otherwise stated."This change is acceptable because, while worded differently, both the CTSand ITS state that ACTIONS do not have to be completed once the LCO is met or is no longer applicable. ITS LCO 3.0.2 also adds the phrase "unlessotherwise stated." There are some ITS ACTIONS that must be completed, DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 3 of 10 even if the LCO is met or is no longer applicable. This change is acceptable because it reflects a new feature in the ITS which does not exist in the CTS. The technical aspects of these changes are discussed in the appropriate ITS sections. These changes are designated as administrative because they are editorial and do not result in technical changes to the Technical Specifications. A04 CTS 3.0.3, in part, is applicable "When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements." ITS LCO 3.0.3 expands those applicability requirements so that the requirement is applicable "When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS." This changes the CTS to add two new applicability conditions.

ITS LCO 3.0.3 is applicable when the LCO is not met and there is no applicable ACTION to be taken. This change is acceptable because it is consistent with the current understanding and application of CTS 3.0.3.
ITS LCO 3.0.3 is applicable when directed by the associated ACTIONS. The CTS and the ITS contain such requirements. Any technical changes related to directing LCO 3.0.3 entry in an ACTION will be discussed in the affected Technical Specifications. This change is acceptable because it is consistent with the current understanding and application of CTS 3.0.3. These changes are designated as administrative because they do not result in any technical changes to the Technical Specifications.

A05 CTS 3.0.3, in part, states that within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in: Hot Standby within the next 6 hours, Hot Shutdown within the following 6 hours, and at least Cold Shutdown within the subsequent 24 hours. ITS LCO 3.0.3 states that action shall be initiated within 1 hour to place the unit, as applicable, in MODE 3 within 7 hours, MODE 4 within 13 hours, and MODE 5 within 37 hours. This changes the CTS by using the sum of the times (i.e., the ITS Completion Time of 37 hours to enter MODE 5 is the same as the sum of the CTS allowance of 1 hour, 6 hours, 6 hours, and 24 hours) instead of sequential times (i.e., each time is measured from the completion of the previous step). The stated times in CTS 3.0.3 and ITS 3.0.3 are listed below: DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 4 of 10 Mode Title CTS Time to Enter Mode ITS Time to Enter Mode -- (Current Mode) 1 hour to begin action 1 hour to begin action 3 Hot Standby within the next 6 hours 7 hours 4 Hot Shutdown within the following 6 hours 13 hours 5 Cold Shutdown within the subsequent 24 hours 37 hours The purpose of CTS 3.0.3 is to establish the shutdown requirements that must be implemented when an LCO is not met and the condition is not specifically addressed in the associated ACTION requirements. The delineated time limit allows the unit to be placed in a safe shutdown MODE when the plant cannot be maintained within the limits for safe operation. The time limit, specified in CTS 3.0.3 to reach the lower MODES of operation, permits the shutdown to proceed in a controlled manner that is well within the specified maximum cooldown rate. Furthermore, the time limit is within the cooldown capabilities of the plant assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that could challenge safety systems under conditions for which this specification applies. In the CTS, this is accomplished by allowing a total of 37 hours for the plant to be in Cold Shutdown when a shutdown is required during the MODE of Operation. In the absence of specific guidance within the CTS, current SQN practice if the unit is in a lower MODE of Operation and a CTS 3.0.3 shutdown is required, is to apply the time limit for reaching the lower MODE of operation (i.e., each time limit is measured from the time the previous MODE is reached). In the ITS, the time limits for ITS LCO 3.0.3 allow 37 hours for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation. If the unit is in a lower mode of operation when an ITS LCO 3.0.3 shutdown is required, the time limit for reaching the next lower MODE applies (i.e., if the plant is in MODE 3, 13 hours is allowed to reach MODE 4). ITS 3.0 Bases gives a detailed discussion on the use of applying the allowed outage times when the unit is in a lower MODE when ITS 3.0.3 is entered. This is further explained, with examples, in the discussion of Section 1.3, "Completion Times." This change is acceptable because ITS and CTS both allow 37 hours to reach MODE 5 from power operation. In addition, the CTS 3.0.3 statement " within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply" has been editorially reworded in ITS LCO 3.0.3 to "the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. ACTION shall be initiated within 1 hour to place the unit..." These changes are considered changes to the CTS presentation. These changes are designated as administrative as they apply rules of usage established by ITS without resulting in technical changes to the Technical Specifications. A06 CTS 3.0.3 states "Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 5 of 10 accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation." ITS LCO 3.0.3 states "Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is applicable in MODES 1, 2, 3, and 4." This change is acceptable because the changes to CTS 3.0.3 are editorial. Both the CTS and ITS state that LCO 3.0.3 can be exited if the LCO which led to the entry into LCO 3.0.3 is met, or if one of the ACTIONS of that LCO is applicable. The CTS requirement also specifies that the time to complete the ACTIONS in the LCO is based on the initial failure to meet the LCO. Reentering the LCO after exiting LCO 3.0.3 does not reset the ACTION statement time requirements. This information is not explicitly stated in ITS LCO 3.0.3 but is true under the multiple condition entry concept of the ITS. In addition, the sentence "LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4" is added to ITS LCO 3.0.3. CTS 3.0.3 and ITS LCO 3.0.3 require the unit to be placed only as low as COLD SHUTDOWN (MODE 5). Once the unit is in MODE 5, there are no further requirements. Thus, CTS 3.0.3 and ITS LCO 3.0.3 are effectively only applicable in MODES 1, 2, 3, and 4, and the addition of the sentence merely reflects editorial preferences used in the ITS. These changes are designated as administrative because there is no change in the intent or application of the CTS 3.0.3 requirements. A07 CTS 3.0.6 has a statement that CTS 3.0.6 is an exception to both CTS 3.0.1 and CTS 3.0.2. ITS LCO 3.0.5 includes only a statement that ITS LCO 3.0.5 is an exception to LCO 3.0.2. The statement that ITS LCO 3.0.5 is an exception to LCO 3.0.1 is not included. This change is acceptable since ITS LCO 3.0.5 does not modify ITS LCO 3.0.1. The ACTION requirements discussion that is in CTS 3.0.1 has been moved to ITS LCO 3.0.2 (i.e., it is not included in ITS LCO 3.0.1). This change is designated as administrative since it does not result in any technical change to the Technical Specifications. A08 ITS LCO 3.0.6 is added to the CTS to provide guidance regarding the appropriate ACTIONS to be taken when a single inoperability (a support system) also results in the inoperability of one or more related systems (supported system(s)). ITS LCO 3.0.6 states "When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 6 of 10 accordance with LCO 3.0.2." In the CTS, based on the intent and interpretation provided by the NRC over the years, there has been an ambiguous approach to the combined support/supported inoperability. Some of this history is summarized below: *Guidance provided in the June 13, 1979, NRC memorandum from Brian K.Grimes (Assistant Director for Engineering and Projects) to Samuel E. Bryan(Assistant Director for Field Coordination) would indicate anintent/interpretation consistent with the proposed LCO 3.0.6, without thenecessity of also requiring additional ACTIONS. That is, only the inoperable support system ACTIONS need be taken.*Guidance provided by the NRC in their April 10, 1980, letter to all Licensees,regarding the definition of OPERABILITY and its impact as a support system on the remainder of the CTS, would indicate a similar philosophy of not takingACTIONS for the inoperable supported equipment. However, in this case,additional actions (similar to the proposed Safety Function Determination Program actions) were addressed and required.*Regulatory Issue Summary (RIS) 2005-20 and a reading of the CTS providean interpretation that inoperability, even as a result of a Technical Specification support system inoperability, requires all associated ACTIONS to be taken.*Certain CTS contain ACTIONS such as "Declare the {supported system}inoperable and take the ACTIONS of {its Specification}." In many cases, the supported system would likely already be considered inoperable. The implication of this presentation is that the ACTIONS of the inoperable supported system would not have been taken without the specific direction to do so.Considering the history of misunderstandings in this area, the WOG ISTS, NUREG-1431, Rev. 4, was developed with Industry input and approval of the NRC to include LCO 3.0.6 and a new program, Specification 5.5.13, "Safety Function Determination Program (SFDP)." This change is acceptable since its function is to clarify existing ambiguities and to maintain actions within the realm of previous interpretations. This change is designated as administrative because it does not technically change the Technical Specifications. A09 ITS LCO 3.0.7 is added to the CTS. ITS LCO 3.0.7 states "Test Exception LCOs 3.1.8, "PHYSICS TEST Exceptions - MODE 2" allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications." DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 7 of 10 This change is acceptable because the CTS contain test exception specifications that allow certain LCOs to not be met for the purpose of special tests and operations. However, the CTS does not contain the equivalent of ITS LCO 3.0.7. As a result, there could be confusion regarding which LCOs are applicable during special tests. LCO 3.0.7 was crafted to avoid that possible confusion. LCO 3.0.7 is consistent with the use and application of CTS test exception Specifications and does not provide any new restriction or allowance. This change is designated as administrative because it does not technically change the Technical Specifications. A10 CTS 4.0.2 states, "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." ITS SR 3.0.2 states, "The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications." This results in several changes to the CTS.

ITS SR 3.0.2 adds to the CTS "For Frequencies specified as "once," the above interval extension does not apply." This is described in DOC M01.
ITS SR 3.0.2 adds to the CTS "If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance." This is covered by DOC L02.
CTS 4.0.2 states, "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." ITS SR 3.0.2 states, in part, "The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency." This change is being made to be consistent with the ITS terminology and to clarify the concept of the specified SR Frequency being met. This change is acceptable since it does not change the intent of the requirements.
ITS SR 3.0.2 is more specific regarding the state of the Frequency by stating, "as measured from the previous performance or as measured from the time a specified condition of the Frequency is met." This direction is consistent with the current use and application of the Technical Specifications.

This change is acceptable because the ITS intent is the same as the CTS requirement. DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 8 of 10 *ITS SR 3.0.2 adds to the CTS "Exceptions to this Specification are stated inthe individual Specifications."This change is acceptable because it reflects practices used in the ITS thatare not used in the CTS. Any changes to a Technical Specification, by inclusion of such an exception, will be addressed in the affected Technical Specification.The changes, except as discussed in DOC M01 and DOC L02, are designated as administrative because they reflect presentation and usage rules of the ITS without making technical changes to the Technical Specifications. A11 CTS 4.0.3 states, in part, "If it is discovered that a Surveillance was not performed within its specified surveillance interval (including the allowed extension per Specification 4.0.2), then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater." ITS SR 3.0.3 contains a similar requirement, but excludes the statement "(including the allowed extension per Specification 4.0.2)." This changes the CTS by not including the statement "(including the allowed extension per Specification 4.0.2)." This change is acceptable because the statement in CTS 4.0.2 "(including the allowed extension per Specification 4.0.2)" is not needed to be repeated in ITS SR 3.0.3. ITS SR 3.0.2 allows a Surveillance to be performed within 1.25 times the interval specified in the Frequency. Therefore, there is no need to repeat in ITS SR 3.0.3 the allowance that is granted in ITS SR 3.0.2. This change is designated as administrative because it does not result in a technical change to the Technical Specifications. MORE RESTRICTIVE CHANGES M01 CTS 4.0.2 states, "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." ITS SR 3.0.2 states, "The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications." This changes the CTS by adding "For Frequencies specified as "once," the above interval extension does not apply." The remaining changes to CTS 4.0.2 are discussed in DOC A10 and DOC L02. The purpose of the 1.25 extension allowance to Surveillance Frequencies is to allow for flexibility in scheduling tests. This change is acceptable because Frequencies specified as "once" are typically condition-based Surveillances in DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 9 of 10 which the first performance demonstrates the acceptability of the current condition. Such demonstrations should be accomplished within the specified Frequency without extension in order to avoid operation in unacceptable conditions. This change is designated as more restrictive because an allowance to extend Frequencies by 25 percent is eliminated for some Surveillances. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 CTS Section 3.0 does not contain an allowance when barriers cannot support their support function. The proposed change to CTS 3.0, "LCO Applicability" adds a new LCO 3.0.9. The addition of LCO 3.0.9 to the CTS is to address barriers which cannot perform their related support function for Technical Specification systems. ITS LCO 3.0.9 allows barriers to be able to not perform their safety function for up to 30 days before declaring the supported system inoperable. Furthermore, due to this addition, an allowance is also needed in LCO 3.0.1. This allowance has been added. Barriers are defined as doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, which are designed to provide for the performance of the safety function for the Technical Specification system after the occurrence of one or more initiating events. The barrier which cannot perform its related support function will be evaluated and managed under the Maintenance Rule plant configuration control requirement, 10 CFR 50.65(a)(4), and the associated industry guidance (NUMARC 93-01, Revision 3). This provision is applicable whether the barrier is affected due to planned maintenance or due to a discovered condition. Should the risk assessment and risk management actions for a specific plant configuration or emergent condition not support the 30 day allowed time, the Maintenance Rule risk management determined allowed time and actions must be implemented or the supported system's LCO be considered not met. Application of LCO 3.0.9 is dependent on the OPERABILITY of at least one train or subsystem of the supported Technical Specification system and the system's ability to mitigate the consequences of the specified initiating events. However, during the 30 day period allowed by LCO 3.0.9, there exists the possibility that the train or subsystem required to be OPERABLE will unexpectedly become inoperable. Absent any further consideration, this would likely result in both trains of a Technical Specification required system being declared inoperable DISCUSSION OF CHANGES ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 10 of 10 (i.e., the train supported by the barriers to which LCO 3.0.9 was being applied and the emergent condition of the inoperable train). This would likely result in entering LCO 3.0.3 and a rapid plant shutdown. While this scenario is of low likelihood, it is of very high consequence to the licensee and, therefore, should be avoided unless necessary to avoid an actual plant risk. As a result, LCO 3.0.9 contains a provision which addresses the emergent condition of the required OPERABLE train or subsystem becoming inoperable while LCO 3.0.9 is being used. LCO 3.0.9 provides 24 hours to either restore the inoperable train or subsystem or to cease relying on the provisions of LCO 3.0.9 to consider the train or subsystem supported by the affected barrier(s) OPERABLE. This 24 hour period is not based on a generic risk evaluation, as it would be difficult to perform such an analysis in a generic fashion. Rather, plant risk during this 24 hour allowance is managed using the contemporaneous risk assessment and management required by 10 CFR 50.65(a)(4) and recognizes the unquantified advantage to plant safety of avoiding a plant shutdown with the associated transition risk. A risk impact of the 30 day allowance for barriers was performed. All Sequoyah initiating events are located on the table depicted in TSTF-427 OR Sequoyah has evaluated the use of LCO 3.0.9 for a barrier protecting against an initiating event not on the table located in TSTF-427 and calculated the frequency ranges within the ranges in the table so the above analysis is applicable for those initiators. Therefore, LCO 3.0.9 can be utilized when inoperable barriers affect Systems, Structures, or Components (SSCs). L02 CTS 4.0.2 states, "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." ITS SR 3.0.2 states, " The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications." This changes the CTS by adding, " If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance." The remaining changes to CTS 4.0.2 are discussed in DOC A10 and DOC M01. This change is acceptable because the 25 percent Frequency extension given to provide scheduling flexibility for Surveillances is equally applicable to Required Actions that must be performed periodically. The initial performance is excluded because the first performance demonstrates the acceptability of the current condition. Such demonstrations should be accomplished within the specified Completion Time with extension in order to avoid operation in unacceptable conditions. This change is designated as less restrictive because addition time is provided to perform some periodic Required Actions. Insert 1 SQN is adopting TSTF-427, Revision 2, as incorporated in NUREG-1431, Revision 4, with no deviations from Specification LCO 3.0.9. TVA has reviewed the TSTF-427 documentation and the technical justifications presented in the model application safety evaluation prepared by the NRC staff and find that the technical justifications presented are applicable to SQN Units 1 and 2. In addition, TVA will be adopting the LCO 3.0.9 Bases, as indicated in the ITS conversion submittal. The only deviations to the LCO 3.0.9 Bases have been made for clarity and are justified in the Bases Justification for Deviations. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) LCO Applicability 3.0 Westinghouse STS 3.0-1 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY

LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 3 within 7 hours,

b. MODE 4 within 13 hours, and

c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or 3.0.1 3.0.2 3.0.3 ;;22;3.0.4 LCO Applicability 3.0 Westinghouse STS 3.0-2Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 3.0 LCO Applicability LCO 3.0.4 (continued) c.When an allowance is stated in the individual value, parameter, orother Specification.This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. LCO 3.0.7 Test Exception LCOs [3.1.8 and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. 3.0.4 3.0.6 DOC A08 DOC A09 s, "PHYSICS TEST Exceptions - MODE 2," 3134 LCO Applicability 3.0 Westinghouse STS 3.0-3Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 1 3.0 LCO Applicability LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and: a.the snubbers not able to perform their associated supportfunction(s) are associated with only one train or subsystem of amultiple train or subsystem supported system or are associated witha single train or subsystem supported system and are able toperform their associated support function within 72 hours; orb.the snubbers not able to perform their associated supportfunction(s) are associated with more than one train or subsystem ofa multiple train or subsystem supported system and are able toperform their associated support function within 12 hours.At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met. LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events. If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s). At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met. 3.0.7 DOC L01 SR Applicability 3.0 Westinghouse STS 3.0-4 Rev. 4.0 Amendment XXX 1SEQUOYAH UNIT 1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. 4.0.1 4.0.2 4.0.3 4.0.4 SR Applicability 3.0 Westinghouse STS 3.0-5 Rev. 4.0 Amendment XXX 1SEQUOYAH UNIT 1 3.0 SR Applicability

SR 3.0.4 (continued) This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.4 LCO Applicability 3.0 Westinghouse STS 3.0-1 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY

LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 3 within 7 hours,

b. MODE 4 within 13 hours, and

c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or 3.0.1 3.0.2 3.0.3 ;;22;3.0.4 LCO Applicability 3.0 Westinghouse STS 3.0-2 Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 3.0 LCO Applicability

LCO 3.0.4 (continued)

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. LCO 3.0.7 Test Exception LCOs [3.1.8 and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. 3.0.4 3.0.6 DOC A08 DOC A09 s, "PHYSICS TEST Exceptions - MODE 2," 3134 LCO Applicability 3.0 Westinghouse STS 3.0-3Rev. 4.0 CTS 1Amendment XXX SEQUOYAH UNIT 2 3.0 LCO Applicability LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and: a.the snubbers not able to perform their associated supportfunction(s) are associated with only one train or subsystem of amultiple train or subsystem supported system or are associated witha single train or subsystem supported system and are able toperform their associated support function within 72 hours; orb.the snubbers not able to perform their associated supportfunction(s) are associated with more than one train or subsystem ofa multiple train or subsystem supported system and are able toperform their associated support function within 12 hours.At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met. LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events. If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s). At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met. 3.0.7 DOC L01 SR Applicability 3.0 Westinghouse STS 3.0-4 Rev. 4.0 Amendment XXX 1SEQUOYAH UNIT 2 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. 4.0.1 4.0.2 4.0.3 4.0.4 SR Applicability 3.0 Westinghouse STS 3.0-5 Rev. 4.0 Amendment XXX 1SEQUOYAH UNIT 2 3.0 SR Applicability

SR 3.0.4 (continued) This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.4 JUSTIFICATION FOR DEVIATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. Changes were made to use correct punctuation, typographical errors, or to make other corrections consistent with the Writers Guide for Improved Technical Specifications, TSTF-GG-05-01. 3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is changed to reflect the current licensing basis. 4. Changes were made to reflect changes made to the Specification.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) LCO Applicability B 3.0 Westinghouse STS B3.0-1Rev. 4.0 1Revision XXX Sequoyah Unit 1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.9 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification). LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that: a.Completion of the Required Actions within the specified CompletionTimes constitutes compliance with a Specification andb.Completion of the Required Actions is not required when an LCO ismet within the specified Completion Time, unless otherwise specified. There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation. Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications. 2; LCO Applicability B 3.0 Westinghouse STS B3.0-2Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES LCO 3.0.2 (continued) The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits." The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing so limits the time both subsystems/trains of a safety function are inoperable and limits the time conditions exist which may result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed. When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable, and the ACTIONS Condition(s) are entered. LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and: a.An associated Required Action and Completion Time is not met andno other Condition applies orb.The condition of the unit is not specifically addressed by theassociated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately. 2; LCO Applicability B 3.0 Westinghouse STS B3.0-3Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES LCO 3.0.3 (continued) This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times. A unit shutdown required in accordance with LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs: a.The LCO is now met,b.A Condition exists for which the Required Actions have now beenperformed, orc.ACTIONS exist that do not have expired Completion Times. TheseCompletion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited. The time limits of LCO 3.0.3 allow 37 hours for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 5, or other applicable MODE, is not reduced. For example, if MODE 3 is reached in 2 hours, then the time allowed for reaching MODE 4 is the next 11 hours, because the total time for reaching 22;; LCO Applicability B 3.0 Westinghouse STSB3.0-4Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES LCO 3.0.3 (continued) MODE 4 is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed. In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 has an Applicability of "During movement of irradiated fuel assemblies in the fuel storage pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications. LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c. LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. 13 13 13 Spent 1spent spent LCO Applicability B 3.0 Westinghouse STS B 3.0-5 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.4 (continued) LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate. The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities to be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability. LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented. 3 LCO Applicability B 3.0 Westinghouse STSB3.0-6Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES LCO 3.0.4 (continued) The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these systems and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable. LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., [Containment Air Temperature, Containment Pressure, MCPR, Moderator Temperature Coefficient]), and may be applied to other Specifications based on NRC plant specific approval. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability. The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5. and4 LCO Applicability B 3.0 Westinghouse STS B 3.0-7 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.4 (continued) Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification. Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service or b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY. This Specification does not provide time to perform any other preventive or corrective maintenance. An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions and must be reopened to perform the required testing.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system. 2; LCO Applicability B 3.0 Westinghouse STS B 3.0-8 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for supported systems that have a support system LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the unit is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.

When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability. However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Specification 5.5.15, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6. 135 LCO Applicability B 3.0 Westinghouse STS B 3.0-9 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.6 (continued) The following examples use Figure B 3.0-1 to illustrate loss of safety function conditions that may result when a TS support system is inoperable. In this figure, the fifteen systems that comprise Train A are independent and redundant to the fifteen systems that comprise Train B. To correctly use the figure to illustrate the SFDP provisions for a cross train check, the figure establishes a relationship between support and supported systems as follows: the figure shows System 1 as a support system for System 2 and System 3; System 2 as a support system for System 4 and System 5; and System 4 as a support system for System 8 and System 9. Specifically, a loss of safety function may exist when a support system is inoperable and: a. A system redundant to system(s) supported by the inoperable support system is also inoperable (EXAMPLE B 3.0.6-1), b. A system redundant to system(s) in turn supported by the inoperable supported system is also inoperable (EXAMPLE B 3.0.6-2), or

c. A system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable (EXAMPLE B 3.0.6-3). For the following examples, refer to Figure B 3.0-1.

EXAMPLE B 3.0.6-1 If System 2 of Train A is inoperable and System 5 of Train B is inoperable, a loss of safety function exists in Systems 5, 10, and 11. EXAMPLE B 3.0.6-2 If System 2 of Train A is inoperable, and System 11 of Train B is inoperable, a loss of safety function exists in System 11. EXAMPLE B 3.0.6-3 If System 2 of Train A is inoperable, and System 1 of Train B is inoperable, a loss of safety function exists in Systems 2, 4, 5, 8, 9, 10 and

11. If an evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

22;; LCO Applicability B 3.0 Westinghouse STS B 3.0-10 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.6 (continued) Figure B 3.0-1 Configuration of Trains and Systems This loss of safety function does not require the assumption of additional single failures or loss of offsite power. Since operations are being restricted in accordance with the ACTIONS of the support system, any resulting temporary loss of redundancy or single failure protection is taken into account. Similarly, the ACTIONS for inoperable offsite circuit(s) and inoperable diesel generator(s) provide the necessary restriction for cross train inoperabilities. This explicit cross train verification for inoperable AC electrical power sources also acknowledges that supported system(s) are not declared inoperable solely as a result of inoperability of a normal or emergency electrical power source (refer to the definition of OPERABILITY). When loss of safety function is determined to exist, and the SFDP requires entry into the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists, consideration must be given to the specific type of function affected. Where a loss of function is solely due to a single Technical Specification support system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction LCO Applicability B 3.0 Westinghouse STS B 3.0-11 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.6 (continued)

source due to low tank level) the appropriate LCO is the LCO for the support system. The ACTIONS for a support system LCO adequately address the inoperabilities of that system without reliance on entering its supported system LCO. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system. LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions. Test Exception LCOs [3.1.8 and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect. The Applicability of a Test Exception LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Test Exception LCOs is optional. A special operation may be performed either under the provisions of the appropriate Test Exception LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Test Exception LCO, the requirements of the Test Exception LCO shall be followed. LCO 3.0.8 LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS) under licensee control. The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii), and, as such, are appropriate for control by the licensee. 4, "PHYSICS TEST Exceptions - MODE 2," s LCO Applicability B 3.0 Westinghouse STS B 3.0-12 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.8 (continued)

If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported system's LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2. LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. LCO 3.0.8.a allows 72 hours to restore the snubber(s) before declaring the supported system inoperable. The 72 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system. LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours to restore the snubber(s) before declaring the supported system inoperable. The 12 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function. LCO 3.0.8 requires that risk be assessed and managed. Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function. LCO Applicability B 3.0 Westinghouse STS B3.0-13Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES -----------------------------------REVIEWER'S NOTE----------------------------------- Adoption of LCO 3.0.9 requires the licensee to make the following commitments: 1.[LICENSEE] commits to the guidance of NUMARC 93-01,Revision 3, Section 11, which provides guidance and details on theassessment and management of risk during maintenance.2.[LICENSEE] commits to the guidance of NEI 04-08, "Allowance forNon Technical Specification Barrier Degradation on SupportedSystem OPERABILITY (TSTF-427) Industry Implementation Guidance," March 2006. -------------------------------------------------------------------------------------------------- LCO 3.0.9 LCO 3.0.9 establishes conditions under which systems described in the Technical Specifications are considered to remain OPERABLE when required barriers are not capable of providing their related support function(s). Barriers are doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, that support the performance of the safety function of systems described in the Technical Specifications. This LCO states that the supported system is not considered to be inoperable solely due to required barriers not capable of performing their related support function(s) under the described conditions. LCO 3.0.9 allows 30 days before declaring the supported system(s) inoperable and the LCO(s) associated with the supported system(s) not met. A maximum time is placed on each use of this allowance to ensure that as required barriers are found or are otherwise made unavailable, they are restored. However, the allowable duration may be less than the specified maximum time based on the risk assessment. If the allowed time expires and the barriers are unable to perform their related support function(s), the supported system's LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2. This provision does not apply to barriers which support ventilation systems or to fire barriers. The Technical Specifications for ventilation systems provide specific Conditions for inoperable barriers. Fire barriers are addressed by other regulatory requirements and associated plant programs. This provision does not apply to barriers which are not required to support system OPERABILITY (see NRC Regulatory Issue Summary 2001-09, "Control of Hazard Barriers," dated April 2, 2001). 6discovered 333because that LCO Applicability B 3.0 Westinghouse STS B3.0-14Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES LCO 3.0.9 (continued) The provisions of LCO 3.0.9 are justified because of the low risk associated with required barriers not being capable of performing their related support function. This provision is based on consideration of the following initiating event categories: -----------------------------------REVIEWER'S NOTE----------------------------------- LCO 3.0.9 may be expanded to other initiating event categories provided plant-specific analysis demonstrates that the frequency of the additional initiating events is bounded by the generic analysis or if plant-specific approval is obtained from the NRC. -------------------------------------------------------------------------------------------------- Loss of coolant accidents;High energy line breaks;Feedwater line breaks;Internal flooding;External flooding;Turbine missile ejection; andTornado or high wind.The risk impact of the barriers which cannot perform their related support function(s) must be addressed pursuant to the risk assessment and management provision of the Maintenance Rule, 10 CFR 50.65 (a)(4), and the associated implementation guidance, Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." This guidance provides for the consideration of dynamic plant configuration issues, emergent conditions, and other aspects pertinent to plant operation with the barriers unable to perform their related support function(s). These considerations may result in risk management and other compensatory actions being required during the period that barriers are unable to perform their related support function(s). LCO 3.0.9 may be applied to one or more trains or subsystems of a system supported by barriers that cannot provide their related support function(s), provided that risk is assessed and managed (including consideration of the effects on Large Early Release and from external events). If applied concurrently to more than one train or subsystem of a multiple train or subsystem supported system, the barriers supporting each of these trains or subsystems must provide their related support function(s) for different categories of initiating events. For example, LCO 3.0.9 may be applied for up to 30 days for more than one train of a multiple train supported system if the affected barrier for one train 6 LCO Applicability B 3.0 Westinghouse STS B 3.0-15 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

LCO 3.0.9 (continued)

protects against internal flooding and the affected barrier for the other train protects against tornado missiles. In this example, the affected barrier may be the same physical barrier but serve different protection functions for each train. If during the time that LCO 3.0.9 is being used, the required OPERABLE train or subsystem becomes inoperable, it must be restored to OPERABLE status within 24 hours. Otherwise, the train(s) or subsystem(s) supported by barriers that cannot perform their related support function(s) must be declared inoperable and the associated LCOs declared not met. This 24 hour period provides time to respond to emergent conditions that would otherwise likely lead to entry into LCO 3.0.3 and a rapid plant shutdown, which is not justified given the low probability of an initiating event which would require the barrier(s) not capable of performing their related support function(s). During this 24 hour period, the plant risk associated with the existing conditions is assessed and managed in accordance with 10 CFR 50.65(a)(4).

SR Applicability B 3.0 Westinghouse STS B 3.0-16 Rev. 4.0 1Revision XXX Sequoyah Unit 1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO. Surveillances may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified Frequency. Additionally, the definitions related to instrument testing (e.g., CHANNEL CALIBRATION) specify that these tests are performed by means of any series of sequential, overlapping, or total steps. Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when: a.The systems or components are known to be inoperable, althoughstill meeting the SRs; orb.The requirements of the Surveillance(s) are known not to be metbetween required Surveillance performances.Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification. Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition. Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status. SR Applicability B 3.0 Westinghouse STS B 3.0-17 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES SR 3.0.1 (continued) Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. Some examples of this process are: a.Auxiliary feedwater (AFW) pump turbine maintenance duringrefueling that requires testing at steam pressures > 800 psi.However, if other appropriate testing is satisfactorily completed, theAFW System can be considered OPERABLE. This allows startup and other necessary testing to proceed until the plant reaches the steam pressure required to perform the testing.b.High pressure safety injection (HPI) maintenance during shutdownthat requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing. SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ." interval. SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 311SISI greater than33percentpercent SR Applicability B 3.0 Westinghouse STS B 3.0-18 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES SR 3.0.2 (continued) 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. An example of where SR 3.0.2 does not apply is in the Containment Leakage Rate Testing Program. This program establishes testing requirements and Frequencies in accordance with the requirements of regulations. The TS cannot in and of themselves extend a test interval specified in the regulations. As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner. The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified. SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met. This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance. The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. 3percent percent percent percent 333 SR Applicability B 3.0 Westinghouse STS B 3.0-19 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

SR 3.0.3 (continued) When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the licensee's Corrective Action Program.

SR Applicability B 3.0 Westinghouse STS B 3.0-20 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

SR 3.0.3 (continued) If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1. SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability. This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to a Surveillance not being met in accordance with LCO 3.0.4. However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability. However, since the LCO is not met in this

SR Applicability B 3.0 Westinghouse STS B 3.0-21 Rev. 4.0 1Revision XXX Sequoyah Unit 1 BASES

SR 3.0.4 (continued) instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3. The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency. LCO Applicability B 3.0 Westinghouse STS B3.0-1Rev. 4.0 1Revision XXX Sequoyah Unit 2 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.9 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification). LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that: a.Completion of the Required Actions within the specified CompletionTimes constitutes compliance with a Specification andb.Completion of the Required Actions is not required when an LCO ismet within the specified Completion Time, unless otherwise specified. There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation. Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications. 2; LCO Applicability B 3.0 Westinghouse STS B3.0-2Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.2 (continued) The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits." The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing so limits the time both subsystems/trains of a safety function are inoperable and limits the time conditions exist which may result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed. When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable, and the ACTIONS Condition(s) are entered. LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and: a.An associated Required Action and Completion Time is not met andno other Condition applies orb.The condition of the unit is not specifically addressed by theassociated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately. 2; LCO Applicability B 3.0 Westinghouse STS B3.0-3Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.3 (continued) This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times. A unit shutdown required in accordance with LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs: a.The LCO is now met,b.A Condition exists for which the Required Actions have now beenperformed, orc.ACTIONS exist that do not have expired Completion Times. TheseCompletion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited. The time limits of LCO 3.0.3 allow 37 hours for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 5, or other applicable MODE, is not reduced. For example, if MODE 3 is reached in 2 hours, then the time allowed for reaching MODE 4 is the next 11 hours, because the total time for reaching 22;; LCO Applicability B 3.0 Westinghouse STSB3.0-4Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.3 (continued) MODE 4 is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed. In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.15, "Fuel Storage Pool Water Level." LCO 3.7.15 has an Applicability of "During movement of irradiated fuel assemblies in the fuel storage pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.15 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.15 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications. LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c. LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. 13 13 13 Spent 1spent spent LCO Applicability B 3.0 Westinghouse STS B3.0-5Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.4 (continued) LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate. The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities to be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability. LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components. The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented. 3 LCO Applicability B 3.0 Westinghouse STSB3.0-6Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.4 (continued) The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these systems and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable. LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., [Containment Air Temperature, Containment Pressure, MCPR, Moderator Temperature Coefficient]), and may be applied to other Specifications based on NRC plant specific approval. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability. The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5. and4 LCO Applicability B 3.0 Westinghouse STS B 3.0-7 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

LCO 3.0.4 (continued) Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification. Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service or b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY. This Specification does not provide time to perform any other preventive or corrective maintenance. An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions and must be reopened to perform the required testing.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system. 2; LCO Applicability B 3.0 Westinghouse STSB3.0-8Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for supported systems that have a support system LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the unit is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions. When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability. However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions. However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. Specification 5.5.15, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6. 135 LCO Applicability B 3.0 Westinghouse STS B3.0-9Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.6 (continued) The following examples use Figure B 3.0-1 to illustrate loss of safety function conditions that may result when a TS support system is inoperable. In this figure, the fifteen systems that comprise Train A are independent and redundant to the fifteen systems that comprise Train B. To correctly use the figure to illustrate the SFDP provisions for a cross train check, the figure establishes a relationship between support and supported systems as follows: the figure shows System 1 as a support system for System 2 and System 3; System 2 as a support system for System 4 and System 5; and System 4 as a support system for System 8 and System 9. Specifically, a loss of safety function may exist when a support system is inoperable and: a.A system redundant to system(s) supported by the inoperablesupport system is also inoperable (EXAMPLE B 3.0.6-1),b.A system redundant to system(s) in turn supported by the inoperablesupported system is also inoperable (EXAMPLE B 3.0.6-2), or c.A system redundant to support system(s) for the supported systems(a) and (b) above is also inoperable (EXAMPLE B 3.0.6-3).For the following examples, refer to Figure B 3.0-1. EXAMPLE B 3.0.6-1 If System 2 of Train A is inoperable and System 5 of Train B is inoperable, a loss of safety function exists in Systems 5, 10, and 11. EXAMPLE B 3.0.6-2 If System 2 of Train A is inoperable, and System 11 of Train B is inoperable, a loss of safety function exists in System 11. EXAMPLE B 3.0.6-3 If System 2 of Train A is inoperable, and System 1 of Train B is inoperable, a loss of safety function exists in Systems 2, 4, 5, 8, 9, 10 and

11. If an evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 22;;

LCO Applicability B 3.0 Westinghouse STS B3.0-10Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.6 (continued) Figure B 3.0-1 Configuration of Trains and Systems This loss of safety function does not require the assumption of additional single failures or loss of offsite power. Since operations are being restricted in accordance with the ACTIONS of the support system, any resulting temporary loss of redundancy or single failure protection is taken into account. Similarly, the ACTIONS for inoperable offsite circuit(s) and inoperable diesel generator(s) provide the necessary restriction for cross train inoperabilities. This explicit cross train verification for inoperable AC electrical power sources also acknowledges that supported system(s) are not declared inoperable solely as a result of inoperability of a normal or emergency electrical power source (refer to the definition of OPERABILITY). When loss of safety function is determined to exist, and the SFDP requires entry into the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists, consideration must be given to the specific type of function affected. Where a loss of function is solely due to a single Technical Specification support system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction LCO Applicability B 3.0 Westinghouse STS B 3.0-11 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

LCO 3.0.6 (continued)

source due to low tank level) the appropriate LCO is the LCO for the support system. The ACTIONS for a support system LCO adequately address the inoperabilities of that system without reliance on entering its supported system LCO. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system. LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions. Test Exception LCOs [3.1.8 and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect. The Applicability of a Test Exception LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Test Exception LCOs is optional. A special operation may be performed either under the provisions of the appropriate Test Exception LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Test Exception LCO, the requirements of the Test Exception LCO shall be followed. LCO 3.0.8 LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS) under licensee control. The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii), and, as such, are appropriate for control by the licensee. 4, "PHYSICS TEST Exceptions - MODE 2," s LCO Applicability B 3.0 Westinghouse STS B3.0-12Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.8 (continued) If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported system's LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2. LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. LCO 3.0.8.a allows 72 hours to restore the snubber(s) before declaring the supported system inoperable. The 72 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system. LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours to restore the snubber(s) before declaring the supported system inoperable. The 12 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function. LCO 3.0.8 requires that risk be assessed and managed. Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function. LCO Applicability B 3.0 Westinghouse STS B3.0-13Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES -----------------------------------REVIEWER'S NOTE----------------------------------- Adoption of LCO 3.0.9 requires the licensee to make the following commitments: 1.[LICENSEE] commits to the guidance of NUMARC 93-01,Revision 3, Section 11, which provides guidance and details on theassessment and management of risk during maintenance.2.[LICENSEE] commits to the guidance of NEI 04-08, "Allowance forNon Technical Specification Barrier Degradation on SupportedSystem OPERABILITY (TSTF-427) Industry Implementation Guidance," March 2006. -------------------------------------------------------------------------------------------------- LCO 3.0.9 LCO 3.0.9 establishes conditions under which systems described in the Technical Specifications are considered to remain OPERABLE when required barriers are not capable of providing their related support function(s). Barriers are doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, that support the performance of the safety function of systems described in the Technical Specifications. This LCO states that the supported system is not considered to be inoperable solely due to required barriers not capable of performing their related support function(s) under the described conditions. LCO 3.0.9 allows 30 days before declaring the supported system(s) inoperable and the LCO(s) associated with the supported system(s) not met. A maximum time is placed on each use of this allowance to ensure that as required barriers are found or are otherwise made unavailable, they are restored. However, the allowable duration may be less than the specified maximum time based on the risk assessment. If the allowed time expires and the barriers are unable to perform their related support function(s), the supported system's LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2. This provision does not apply to barriers which support ventilation systems or to fire barriers. The Technical Specifications for ventilation systems provide specific Conditions for inoperable barriers. Fire barriers are addressed by other regulatory requirements and associated plant programs. This provision does not apply to barriers which are not required to support system OPERABILITY (see NRC Regulatory Issue Summary 2001-09, "Control of Hazard Barriers," dated April 2, 2001). 6discovered 333because that LCO Applicability B 3.0 Westinghouse STS B3.0-14Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES LCO 3.0.9 (continued) The provisions of LCO 3.0.9 are justified because of the low risk associated with required barriers not being capable of performing their related support function. This provision is based on consideration of the following initiating event categories: -----------------------------------REVIEWER'S NOTE----------------------------------- LCO 3.0.9 may be expanded to other initiating event categories provided plant-specific analysis demonstrates that the frequency of the additional initiating events is bounded by the generic analysis or if plant-specific approval is obtained from the NRC. -------------------------------------------------------------------------------------------------- *Loss of coolant accidents;*High energy line breaks;*Feedwater line breaks;*Internal flooding;*External flooding;*Turbine missile ejection; and*Tornado or high wind.The risk impact of the barriers which cannot perform their related support function(s) must be addressed pursuant to the risk assessment and management provision of the Maintenance Rule, 10 CFR 50.65 (a)(4), and the associated implementation guidance, Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." This guidance provides for the consideration of dynamic plant configuration issues, emergent conditions, and other aspects pertinent to plant operation with the barriers unable to perform their related support function(s). These considerations may result in risk management and other compensatory actions being required during the period that barriers are unable to perform their related support function(s). LCO 3.0.9 may be applied to one or more trains or subsystems of a system supported by barriers that cannot provide their related support function(s), provided that risk is assessed and managed (including consideration of the effects on Large Early Release and from external events). If applied concurrently to more than one train or subsystem of a multiple train or subsystem supported system, the barriers supporting each of these trains or subsystems must provide their related support function(s) for different categories of initiating events. For example, LCO 3.0.9 may be applied for up to 30 days for more than one train of a multiple train supported system if the affected barrier for one train 6 LCO Applicability B 3.0 Westinghouse STS B 3.0-15 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

LCO 3.0.9 (continued)

protects against internal flooding and the affected barrier for the other train protects against tornado missiles. In this example, the affected barrier may be the same physical barrier but serve different protection functions for each train. If during the time that LCO 3.0.9 is being used, the required OPERABLE train or subsystem becomes inoperable, it must be restored to OPERABLE status within 24 hours. Otherwise, the train(s) or subsystem(s) supported by barriers that cannot perform their related support function(s) must be declared inoperable and the associated LCOs declared not met. This 24 hour period provides time to respond to emergent conditions that would otherwise likely lead to entry into LCO 3.0.3 and a rapid plant shutdown, which is not justified given the low probability of an initiating event which would require the barrier(s) not capable of performing their related support function(s). During this 24 hour period, the plant risk associated with the existing conditions is assessed and managed in accordance with 10 CFR 50.65(a)(4).

SR Applicability B 3.0 Westinghouse STS B 3.0-16 Rev. 4.0 1Revision XXX Sequoyah Unit 2 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO. Surveillances may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified Frequency. Additionally, the definitions related to instrument testing (e.g., CHANNEL CALIBRATION) specify that these tests are performed by means of any series of sequential, overlapping, or total steps. Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when: a.The systems or components are known to be inoperable, althoughstill meeting the SRs; orb.The requirements of the Surveillance(s) are known not to be metbetween required Surveillance performances.Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification. Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition. Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status. SR Applicability B 3.0 Westinghouse STS B 3.0-17 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES SR 3.0.1 (continued) Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. Some examples of this process are: a.Auxiliary feedwater (AFW) pump turbine maintenance duringrefueling that requires testing at steam pressures > 800 psi.However, if other appropriate testing is satisfactorily completed, theAFW System can be considered OPERABLE. This allows startup and other necessary testing to proceed until the plant reaches the steam pressure required to perform the testing.b.High pressure safety injection (HPI) maintenance during shutdownthat requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing. SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ." interval. SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 311SISI greater than33percentpercent SR Applicability B 3.0 Westinghouse STS B 3.0-18 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES SR 3.0.2 (continued) 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. An example of where SR 3.0.2 does not apply is in the Containment Leakage Rate Testing Program. This program establishes testing requirements and Frequencies in accordance with the requirements of regulations. The TS cannot in and of themselves extend a test interval specified in the regulations. As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner. The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified. SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met. This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance. The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. 3percent percent percent percent 333 SR Applicability B 3.0 Westinghouse STS B 3.0-19 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

SR 3.0.3 (continued) When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the licensee's Corrective Action Program.

SR Applicability B 3.0 Westinghouse STS B 3.0-20 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

SR 3.0.3 (continued) If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1. SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability. This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to a Surveillance not being met in accordance with LCO 3.0.4. However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability. However, since the LCO is not met in this

SR Applicability B 3.0 Westinghouse STS B 3.0-21 Rev. 4.0 1Revision XXX Sequoyah Unit 2 BASES

SR 3.0.4 (continued) instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3. The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency. JUSTIFICATION FOR DEVIATIONS ITS 3.0 BASES, LCO AND SR APPLICABILITY Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS Bases whichreflect the plant specific nomenclature, number, reference, system description,analysis, or licensing basis description.2.These punctuation corrections have been made consistent with the Writer's Guidefor the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.3.Changes have been made for clarity.4.The ISTS contains bracketed information and/or values that are generic to allWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is changed to reflect the current licensing basis.5.Changes have been made to reflect changes made to the Specification.6.The Reviewer's Note has been deleted. This information is for the NRC reviewer tobe keyed into what is needed to meet this requirement. This Note is not meant to beretained in the final version of the plant specific submittal. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 1 of 5 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L01 The Tennessee Valley Authority (TVA) is converting Sequoyah to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, Rev. 4, "Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the Current Technical Specifications (CTS) Less Restrictive. Below is the description of this less restrictive change and the determination of No Significant Hazards Considerations for conversion to NUREG-1431. CTS Section 3.0 does not contain an allowance when barriers cannot support their support function. The proposed change to CTS 3.0, "LCO Applicability" adds a new LCO 3.0.9. The addition of LCO 3.0.9 to the CTS is to address barriers which cannot perform their related support function for Technical Specification systems. ITS LCO 3.0.9 allows barriers to be able to not perform their safety function for up to 30 days before declaring the supported system inoperable. Furthermore, due to this addition, an allowance is also needed in LCO 3.0.1. This allowance has been added.

Barriers are defined as doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, which are designed to provide for the performance of the safety function for the Technical Specification system after the occurrence of one or more initiating events. The barrier which cannot perform its related support function will be evaluated and managed under the Maintenance Rule plant configuration control requirement, 10 CFR 50.65(a)(4), and the associated industry guidance (NUMARC 93-01, Revision 3). This provision is applicable whether the barrier is affected due to planned maintenance or due to a discovered condition. Should the risk assessment and risk management actions for a specific plant configuration or emergent condition not support the 30 day allowed time, the Maintenance Rule risk management determined allowed time and actions must be implemented or the supported system's LCO be considered not met. Application of LCO 3.0.9 is dependent on the OPERABILITY of at least one train or subsystem of the supported Technical Specification system and the system's ability to mitigate the consequences of the specified initiating events. However, during the 30 day period allowed by LCO 3.0.9, there exists the possibility that the train or subsystem required to be OPERABLE will unexpectedly become inoperable. Absent any further consideration, this would likely result in both trains of a Technical Specification required system being declared inoperable (i.e., the train supported by the barriers to which LCO 3.0.9 was being applied and the emergent condition of the inoperable train). This would likely result in entering LCO 3.0.3 and a plant shutdown. While this scenario is of low likelihood, it is of very high consequence to the licensee and, therefore, should be avoided unless necessary to avoid an actual plant risk. As a result, LCO 3.0.9 contains a provision which addresses the emergent condition of the required OPERABLE train or subsystem becoming inoperable while LCO 3.0.9 is being used. LCO 3.0.9 provides 24 hours to either restore the inoperable train or subsystem or to cease relying on the provisions of LCO 3.0.9 to consider the train or subsystem supported by the affected barrier(s) OPERABLE. This 24 hour period is not based on a generic risk evaluation, as it would be difficult to perform such an analysis in a generic fashion. Rather, plant risk DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 2 of 5 during this 24 hour allowance is managed using the contemporaneous risk assessment and management required by 10 CFR 50.65(a)(4) and recognizes the unquantified advantage to plant safety of avoiding a plant shutdown with the associated transition risk.

A risk impact of the 30 day allowance for barriers was performed. All Sequoyah initiating events are located on the table depicted in TSTF-427 OR Sequoyah has evaluated the use of LCO 3.0.9 for a barrier protecting against an initiating event not on the table located in TSTF-427 and calculated the frequency ranges within the ranges in the table so the above analysis is applicable for those initiators. Therefore, LCO 3.0.9 can be utilized when inoperable barriers affect Systems, Structures, or Components (SSCs). TVA has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Barriers are not an initiator to any accident previously evaluated. The probability of an accident previously evaluated is not significantly increased. Barriers support the operation of equipment assumed to mitigate the effects of accidents previously evaluated. The proposed relaxation may only be applied to a single train or subsystem of a multiple train or subsystem Technical Specification system at a given time for a given category of initiating event, or to multiple trains or subsystems of a multiple train or subsystem Technical Specification system provided the affected barriers protect against different categories of initiating events. Therefore, for any given category of initiating event, the ability to perform the assumed safety function is preserved. The consequences of an accident occurring during the time allowed when barriers are not capable of performing their related support function are no different from the consequences of the same accident while relying on the Actions of the supported Technical Specification systems. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new or different accidents result from using the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 3 of 5 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed change allows for a limited period of time in which barriers may be unable to perform their related support function without declaring the supported systems inoperable. A risk analysis has shown that this provision will not have a significant effect on plant risk. In addition, regulatory requirements in 10 CFR 50.65(a)(4) require risk assessment and risk management, which will ensure that plant risk is not significantly increased. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, TVA concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 4 of 5 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L02 The Tennessee Valley Authority (TVA) is converting Sequoyah to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, Rev. 4, "Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the Current Technical Specifications (CTS) Less Restrictive. Below is the description of this less restrictive change and the determination of No Significant Hazards Considerations for conversion to NUREG-1431. CTS 4.0.2 states, "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." ITS SR 3.0.2 states, " The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications." This changes the CTS by adding, " If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance." The remaining changes to CTS 4.0.2 are discussed in DOC A10 and DOC M01.

This change is acceptable because the 25 percent Frequency extension given to provide scheduling flexibility for Surveillances is equally applicable to Required Actions that must be performed periodically. The initial performance is excluded because the first performance demonstrates the acceptability of the current condition. Such demonstrations should be accomplished within the specified Completion Time with extension in order to avoid operation in unacceptable conditions. This change is designated as less restrictive because addition time is provided to perform some periodic Required Actions. TVA has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change allows the Completion Time for periodic actions to be extended by 25 percent. This change does not significantly affect the probability of an accident. The length of time between performance of Required Actions is not an initiator to any accident previously evaluated. The consequences of any accident previously evaluated are the same during the Completion Time or during any extension of the Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0, LCO AND SR APPLICABILITY Sequoyah Unit 1 and 2 Page 5 of 5 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change allows the Completion Time for periodic actions to be extended by 25 percent. This change will not involve physically altering the plant (i.e., no new or different type of equipment will be installed). In addition, the change does not involve any new or revised operator actions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed change allows the Completion Time for periodic actions to be extended by 25 percent. The 25 percent extension allowance is provided for scheduling convenience and is not expected to have significant effect on the average time between Required Actions. As a result, the Required Action will continue to provide appropriate compensatory measures for the subject Condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, TVA concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified. ENCLOSURE 2 VOLUME 6 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.1 REACTIVITY CONTROL SYSTEMS Revision 0 LIST OF ATTACHMENTS 1. ITS Section 3.1.1 - Shutdown Margin 2. ITS Section 3.1.2 - Core Reactivity

3. ITS Section 3.1.3 - Moderator Temperature Coefficient (MTC) 4. ITS Section 3.1.4 - Rod Group Alignment Limits

5. ITS Section 3.1.5 - Shutdown Bank Insertion Limits 6. ITS Section 3.1.6 - Control Bank Insertion Limits 7. ITS Section 3.1.7 - Rod Position Indication 8. ITS Section 3.1.8 - Physics Test Exceptions - MODE 2

9. Relocated/Deleted Current Technical Specifications (CTS)

ATTACHMENT 1 ITS 3.1.1, SHUTDOWN MARGIN (SDM) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS A01ITS 3.1.1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg Greater Than 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s). b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

____________________ *See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-1 Amendment No. 172 Page 1 of 6 A01A02LA01within the limits specified in the COLRLA01not within limitskeff < 1.0 within 15 minutesL01A03A04L02LA01within the limits specified in the COLR See ITS 3.1.4 See ITS Chapter 1.0See ITS 3.1.6 A04LCO 3.1.1 Applicability ACTION A SR 3.1.1.1 See ITS 3.1.6 ITS A01ITS 3.1.1 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6. e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

SEQUOYAH - UNIT 1 3/4 1-2 Page 2 of 6 L03LA03See ITS 3.1.2 MODE 2 with keff < 1.0 M01SR 3.1.1.1 In accordance with the Surveillance Frequency Control Program LA02 ITS A01ITS 3.1.1 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration, 2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-3 Amendment No. 12, 172 Page 3 of 6 A02LA01within the limits specified in the COLRLA01not within limitswithin 15 minutesL01LA01within the limits specified in the COLR LA03LCO 3.1.1 Applicability ACTION A SR 3.1.1.1 L02See ITS 3.1.4 See ITS Chapter 1.0In accordance with the Surveillance Frequency Control Program LA02 A01ITS 3.1.1 ITS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2, with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-1 Amendment No. 163 Page 4 of 6 A01A02LA01within the limits specified in the COLR LA01not within limitskeff < 1.0within 15 minutesL01A03A04L02LA01within the limits specified in the COLR See ITS 3.1.4 See ITS Chapter 1.0See ITS 3.1.6 A04LCO 3.1.1 Applicability ACTION A SR 3.1.1.1 L03See ITS 3.1.6 A01ITS 3.1.1 ITS REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

SEQUOYAH - UNIT 2 3/4 1-2 Page 5 of 6 LA03See ITS 3.1.2 MODE 2 with keff < 1.0 M01SR 3.1.1.1 In accordance with the Surveillance Frequency Control Program LA02 A01ITS 3.1.1 ITS REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration, 2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-3 Amendment No. 163 Page 6 of 6 A02LA01within the limits specified in the COLRLA01within 15 minutesL01LA01within the limits specified in the COLR LA03LCO 3.1.1 Applicability ACTION A SR 3.1.1.1 L02See ITS 3.1.4 See ITS Chapter 1.0In accordance with the Surveillance Frequency Control Program LA02not within limits DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 1 of 5 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.1.1.1 provides the SHUTDOWN MARGIN (SDM) requirement in MODES 1, 2, 3, and 4 (i.e., Tavg greater than 200°F). CTS 3.1.1.2 provides the SDM requirement in MODE 5 (i.e., Tavg less than or equal to 200°F). ITS 3.1.1 provides the SDM requirement in MODE 2 with keff < 1.0 and MODES 3, 4, and 5. This changes the CTS by combining the SDM requirements in MODE 2 with keff < 1.0 and MODES 3, 4, and 5. The change in Applicability for MODE 2 with keff < 1.0 is described in DOC A03. This change is acceptable because the requirements have not changed. Combining the Specifications is an editorial change. Any technical changes resulting from this combination are discussed in other DOCs. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 3.1.1.1 provides the SDM requirement in MODES 1, 2, 3, and 4 (i.e., Tavg greater than 200°F). CTS 4.1.1.1.1 states, when in MODES 1 and 2 with keff 1.0, verify the control bank withdrawal is within the limits of Specification 3.1.3.6. ITS 3.1.1 is Applicable in MODE 2 with keff < 1.0 and MODES 3, 4, and 5. This changes the CTS by combining the SDM requirement in MODE 2 with keff < 1.0 and MODES 3, 4, and 5. The change in Applicability for MODE 1 and MODE 2 with keff 1.0 is described in ITS 3.1.6 (Control Bank Insertion Limits). The purpose of CTS 3.1.1.1 is to ensure that the SDM assumed in the accident analysis is available. When the reactor is critical, SDM is verified by ensuring the control rods are within the control rod insertion limits. ITS 3.1.1 Applicability Bases state in MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." This change is acceptable because the SDM requirements have not changed. Even though CTS 3.1.1.1 is applicable in MODES 1 and 2, the CTS Surveillances only require the verification that control rod bank withdrawal is within the control rod insertion limits. The ITS verifies SDM in MODES 1 and 2 by the rod insertion limits. Any changes to the rod insertion limit requirements are discussed in DOCs for those Specifications. This change is designated as administrative because it does not result in a technical change to the CTS. A04 CTS 3.1.1.1 Applicability is MODES 1, 2, 3, and 4 with a footnote (footnote *) for MODE 2 stating "See Special Test Exception 3.10.1." ITS 3.1.1 does not contain DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 2 of 5 the footnote or a reference to the Special Test Exception. This changes the CTS by not including footnote

in the ITS.

The purpose of the footnote reference is to alert the user that a Special Test Exception exists that may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 4.1.1.1.1.e requires SDM to be determined to be within its limits every 24 hours when in MODES 3 and 4. ITS SR 3.1.1.1 requires SDM to be determined to be within its limits in MODE 2 with keff < 1.0 and MODES 3 and 4. This changes the CTS by expanding the applicability of the Surveillance to include MODE 2 with keff < 1.0. The purpose of CTS 4.1.1.1.1.e is to verify that sufficient SDM is available. CTS 4.1.1.1.1.b states that when the reactor is in MODE 1 and MODE 2 with keff 1.0, SDM is verified by determining that the control rods are above the rod insertion limits. In MODE 2 with keff < 1.0, CTS 4.1.1.1.1.c verifies SDM by determining that the control rods are above the rod insertion limits. However, no CTS Surveillance requires a periodic verification of SDM when in MODE 2 with keff < 1.0. This change is acceptable because the ITS requires a specific verification that the SDM is within the limit when in MODE 2 with keff < 1.0 on a periodic basis. This change is designated as more restrictive because it expands the conditions under which a Surveillance must be performed. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 6 - Removal of Cycle-Specific Parameter Limits from the Technical Specifications to the Core Operating Limits Report) CTS 3.1.1.1, CTS 3.1.1.1 ACTION and CTS 4.1.1.1.1 require the SDM to be greater than or equal to 1.6% delta k/k when in MODES 1, 2, 3, and 4. CTS 3.1.1.2, CTS 3.1.1.2 ACTION and CTS 4.1.1.2.1 require the SDM to be greater than or equal to 1.0% delta k/k when in MODE 5. ITS LCO 3.1.1 requires the SDM to be within the limits specified in the COLR. ITS 3.1.1 ACTION A provides actions when the SDM is not within limits. ITS SR 3.1.1.1 requires verification that the SDM is within limits. This changes the CTS by moving the SDM limits to the COLR. The removal of these cycle-specific parameter limits from the Technical Specifications to the COLR is acceptable because the cycle-specific limits are developed or utilized under NRC-approved methodologies that will ensure that the safety limits are met. The NRC documented in Generic Letter 88-16, DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 3 of 5 "Removal of Cycle-Specific Parameter Limits From Technical Specifications," that this type of information is not necessary to be included in the Technical Specification to provide adequate protection of public health and safety. The ITS retains the SDM requirement. The methodologies used to develop the parameters in the COLR have obtained approval by the NRC in accordance with Generic Letter 88-16. Furthermore, this change is acceptable because the removed information will be adequately controlled in the COLR under the requirements provided in ITS 5.6.3, "Core Operating Limits Report." ITS 5.6.3 ensures the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System limits, and nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analyses are met. This change is designated as a less restrictive removal of detail change because information relating to cycle-specific parameter limits is being removed from the Technical Specifications. LA02 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.1.1.1.1.e and CTS 4.1.1.2.b require SDM to be determined to be within its limits every 24 hours. ITS SR 3.1.1.1 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.1.1.1.1.e and CTS 4.1.1.2.b require determination that the SDM is within limits, and specifically requires the consideration of the following factors: reactor coolant system boron concentration, control rod position, reactor coolant system average temperature, fuel burnup based on gross thermal energy generation, xenon concentration and samarium concentration. ITS SR 3.1.1.1 requires a determination that the SDM is within limits, but does not describe the factors that must be considered in the calculation. This information is moved to the Bases. This changes the CTS by removing details on how the SDM calculation is performed from the Specification and placing the information in the Bases. DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 4 of 5 The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement that the SDM be within limits. The detail of how SDM is calculated does not need to appear in the specification in order for the requirement to apply. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.1.1.1 ACTION states when the SDM is less than the applicable limit, boration must be initiated immediately. ITS 3.1.1 ACTION states when SDM is not within limits, boration must be initiated within 15 minutes. This changes the CTS by relaxing the Completion Time from "immediately" to 15 minutes. The purpose of CTS 3.1.1.1 ACTION is to restore the SDM to within its limit promptly. This change is acceptable because the Completion Time is consistent with safe operation under the specific Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, and the low probability of a DBA occurring during the allowed Completion Time. This ITS Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. In addition, the ITS Bases for the ACTION states that boration must be initiated promptly. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS. L02 (Category 4 - Relaxation of Required Action) CTS 3.1.1.1 ACTION states when the SDM is less than or equal to 1.6% k/k, boration must be initiated and continued at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SDM is restored. ITS 3.1.1 ACTION A states that when the SDM is not within limits to initiate boration to restore SDM to within limits. This changes the CTS by eliminating the specific values of flow rate and the boron concentration used to restore compliance with the LCO. The purpose of CTS 3.1.1.1 ACTION is to restore the SDM to within its limit. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 5 of 5 specified redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. Removing the specific values of flow rate and boron concentration from the CTS ACTION provides flexibility in the restoration of the SDM and eliminates conflicts between the SDM value and the specific boration values in the CTS ACTION. As stated, in the ITS Bases for ACTION A, "In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions." Specifying a minimum flow rate and concentration in the ACTION may not accomplish the objective of raising the RCS boron concentration as soon as possible. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L03 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.1.1.1.d requires verification that the SDM is within limit, "Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below (CTS 4.1.1.1.1.e), with the control banks at the maximum insertion limit of Specification 3.1.3.6." The ITS does not contain a similar requirement. This changes the CTS by deleting Surveillance Requirement 4.1.1.1.1.d. The purpose of CTS 4.1.1.1.1.d is to verify core design predictions by determining the SDM with the control rods at the insertion limits. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify the LCO is within limit. The core design predictions, such as rod worth, boron worth, and critical boron concentration, are verified in a manner and at a Frequency necessary to give confidence that these predicted values are within limit in accordance with ITS SR 3.1.2.1. ITS SR 3.1.2.1 has a conditional Frequency similar to that of CTS 4.1.1.1.d requiring performance once prior to entering MODE 1 (> 5% RTP) after each refueling. To ensure the SDM is within limits during reactor startup the critical boron concentration is verified during the startup physics test program and prior to criticality per ITS SR 3.1.6.1 (Estimated Critical Position). Thereafter SDM is confirmed by performance of ITS SR 3.1.4.1 (Rod Alignment), SR 3.1.5.1(Shutdown Bank Rod Insertion Limits), and SR 3.1.6.2 (Control Bank Rod Insertion Limits). Thus, the SDM continues to be verified in a manner and at a Frequency necessary to give confidence that the parameter is within limit. Therefore, the core design parameters upon which SDM relies are verified before exceeding 5% RATED THERMAL POWER after each refueling outage. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) SDM 3.1.1 Westinghouse STS 3.1.1-1 Rev. 4.0 2SEQUOYAH UNIT 1 Amendment XXXCTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limits specified in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits.

A.1 Initiate boration to restore SDM to within limits. 15 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within the limits specified in the COLR. [ 24 hours OR In accordance with the Surveillance Frequency Control Program ] 3.1.1.1, 3.1.1.2 3.1.1.1 Applicability, 3.1.1.2 Applicability 3.1.1.1 ACTION, 3.1.1.2 ACTION 4.1.1.1.1.e, 4.1.1.2.b 11 SDM 3.1.1 Westinghouse STS 3.1.1-1 Rev. 4.0 2SEQUOYAH UNIT 2 Amendment XXXCTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limits specified in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits.

A.1 Initiate boration to restore SDM to within limits. 15 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within the limits specified in the COLR. [ 24 hours OR In accordance with the Surveillance Frequency Control Program ] 3.1.1.1, 3.1.1.2 3.1.1.1 Applicability, 3.1.1.2 Applicability 3.1.1.1 ACTION, 3.1.1.2 ACTION 4.1.1.1.1.e, 4.1.1.2.b 11 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1, SHUTDOWN MARGIN Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. ISTS SR 3.1.1.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) SDM B 3.1.1 Westinghouse STS B 3.1.1-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Control Rod System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY analyses. The safety analysis (Ref. 2) establishes an SDM that ensures ANALYSES specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on scram. For MODE 5, the primary safety analysis that relies on the SDM limits is the boron dilution analysis. 2 SDM B 3.1.1 Westinghouse STS B 3.1.1-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

APPLICABLE SAFETY ANALYSES (continued) The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events,

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and 280 cal/gm energy deposition for the rod ejection accident), and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The most limiting accident for the SDM requirements is based on a main steam line break (MSLB), as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1. In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. Inadvertent boron dilution, b. An uncontrolled rod withdrawal from subcritical or low power condition, s12double ended SDM B 3.1.1 Westinghouse STS B 3.1.1-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

APPLICABLE SAFETY ANALYSES (continued)

c. Startup of an inactive reactor coolant pump (RCP), and d. Rod ejection.

Each of these events is discussed below. In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest. Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits. The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of core power.

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions. LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration. an overtemperature T 1 SDM B 3.1.1 Westinghouse STS B 3.1.1-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

LCO (continued) The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable. APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6. ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met. In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid storage tank, or the borated water storage tank. The operator should borate with the best source available for the plant conditions. In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% k/k must be recovered and a boration flow rate of [ ] gpm, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% k/k. These boration parameters of [ ] gpm and [ ] ppm represent typical values and are provided for the purpose of offering a specific example. 3refueling1(1000 pcm) 50INSERT 1 B 3.1.1 Insert Page B 3.1.1-4 INSERT 1 147 ppm in approximately 46 minutes. If a boron worth of 6.8 pcm/ppm is assumed, this combination will increase the SDM by 1% k/k or 1000 pcm. These boration parameters represent Sequoyah typical values and are provided for the purpose of offering a specific example. 1 SDM B 3.1.1 Westinghouse STS B 3.1.1-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with Keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth. In MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration,

b. Control bank position, c. RCS average temperature, d. Fuel burnup based on gross thermal energy generation, e. Xenon concentration, f. Samarium concentration, and g. Isothermal temperature coefficient (ITC). Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

[ The Frequency of 24 hours is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation. OR  The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] 45462MODE 2 with keff < 1.0 and in SDM B 3.1.1 Westinghouse STS B 3.1.1-6 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Chapter [15].

3. FSAR, Chapter [15].

4. 10 CFR 100.

Section 15.4.2 Section 15.2.4 U 1313 SDM B 3.1.1 Westinghouse STS B 3.1.1-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Control Rod System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY analyses. The safety analysis (Ref. 2) establishes an SDM that ensures ANALYSES specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on scram. For MODE 5, the primary safety analysis that relies on the SDM limits is the boron dilution analysis. 2 SDM B 3.1.1 Westinghouse STS B 3.1.1-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

APPLICABLE SAFETY ANALYSES (continued) The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events,

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and 280 cal/gm energy deposition for the rod ejection accident), and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The most limiting accident for the SDM requirements is based on a main steam line break (MSLB), as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1. In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. Inadvertent boron dilution, b. An uncontrolled rod withdrawal from subcritical or low power condition, s12double ended SDM B 3.1.1 Westinghouse STS B 3.1.1-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

APPLICABLE SAFETY ANALYSES (continued)

c. Startup of an inactive reactor coolant pump (RCP), and d. Rod ejection.

Each of these events is discussed below. In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest. Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits. The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of core power.

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions. LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration. an overtemperature T 1 SDM B 3.1.1 Westinghouse STS B 3.1.1-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

LCO (continued) The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable. APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6. ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met. In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid storage tank, or the borated water storage tank. The operator should borate with the best source available for the plant conditions. In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% k/k must be recovered and a boration flow rate of [ ] gpm, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% k/k. These boration parameters of [ ] gpm and [ ] ppm represent typical values and are provided for the purpose of offering a specific example. 3refueling1(1000 pcm) 50INSERT 1 B 3.1.1 Insert Page B 3.1.1-4 INSERT 1 156 ppm in approximately 48 minutes. If a boron worth of 6.4 pcm/ppm is assumed, this combination will increase the SDM by 1% k/k or 1000 pcm. These boration parameters represent Sequoyah typical values and are provided for the purpose of offering a specific example. 1 SDM B 3.1.1 Westinghouse STS B 3.1.1-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with Keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth. In MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration,

b. Control bank position, c. RCS average temperature, d. Fuel burnup based on gross thermal energy generation, e. Xenon concentration, f. Samarium concentration, and g. Isothermal temperature coefficient (ITC). Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

[ The Frequency of 24 hours is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation. OR  The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] 45462MODE 2 with keff < 1.0 and in SDM B 3.1.1 Westinghouse STS B 3.1.1-6 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Chapter [15].

3. FSAR, Chapter [15].

4. 10 CFR 100.

Section 15.4.2 Section 15.2.4 U 1313 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. Editorial changes made for enhanced clarity/consistency.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 4. ISTS SR 3.1.1.1 Bases provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 6. Changes are made to be consistent with the Specification.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.1, SHUTDOWN MARGIN Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 2 ITS 3.1.2, CORE REACTIVITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.2 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg Greater Than 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s). b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

____________________ *See Special Test Exception 3.10.1 November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-1 Amendment No. 172 Page 1 of 4 See ITS 3.1.1 See ITS 3.1.1 Core ReactivityA02See ITS 3.1.1 See ITS 3.1.4 See ITS 3.1.6 See ITS 3.1.1 L02Add proposed ACTIONS A and BApplicability L01A02Add proposed LCO 3.1.2 SR 3.1.2.1 A01ITS ITS 3.1.2 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6. e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

SEQUOYAH - UNIT 1 3/4 1-2 See ITS 3.1.1 SR 3.1.2.1 SR 3.1.2.1 Note LA02LA01In accordance with the Surveillance Frequency Control Program L03Page 2 of 4 may L04SR 3.1.2.1 Prior to entering MODE 1 after refueling and L03LA02 A01ITS ITS 3.1.2 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2, with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-1 Amendment No. 163 Page 3 of 4 See ITS 3.1.1 See ITS 3.1.1 Core ReactivityA02See ITS 3.1.1 See ITS 3.1.6 L02Add proposed ACTIONS A and BApplicability L01A02Add proposed LCO 3.1.2 See ITS 3.1.1 See ITS 3.1.4 See ITS Chapter 1.0SR 3.1.2.1 A01ITS ITS 3.1.2 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration, 2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

SEQUOYAH - UNIT 2 3/4 1-2 See ITS 3.1.1 SR 3.1.2.1 Note LA02LA01In accordance with the Surveillance Frequency Control Program L03Page 4 of 4 may L04SR 3.1.2.1 SR 3.1.2.1 Prior to entering MODE 1 after refueling and L03LA02 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 1 of 5 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 4.1.1.1.2 requires the overall core reactivity balance to be compared to predicted values to demonstrate agreement within +/- 1% k/k. However, this Surveillance is currently part of the SHUTDOWN MARGIN Specification. Additionally, CTS 3.1.1.1 is titled SHUTDOWN MARGIN - Tavg Greater Than 200°F. A new LCO, ITS LCO 3.1.2, requires the measured core reactivity to be within +/- 1% k/k of predicted values. Furthermore, ITS 3.1.2 is titled Core Reactivity. This changes the CTS by having a separate Specification for the Core Reactivity requirement and changing the title. This change is acceptable because the requirements have not changed. Converting the requirement from a Surveillance in the SHUTDOWN MARGIN specification to an LCO is consistent with the ITS format and content guidance. Any technical changes resulting from this change are discussed in other DOCs. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS requires the measured core reactivity to be determined to be within +/- 1% k/k of the predicted value at least every 31 Effective Full Power Days (EFPD). ITS SR 3.1.2.1 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program. DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 2 of 5 The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.1.1.1.2 requires comparison of the actual and predicted core reactivity balance and specifically requires consideration of at least those factors stated in Specification 4.1.1.1.1.e. CTS 4.1.1.1.1.e requires determination of SDM and requires the consideration of the following factors: reactor coolant system boron concentration, control rod position, reactor coolant system average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, and samarium concentration. ITS SR 3.1.2.1 requires comparison of the actual and predicted core reactivity, but does not describe the factors that must be considered in the calculation. This information is relocated to the Bases. This changes the CTS by removing details on how the core reactivity balance comparison calculation is performed from the CTS and placing the information in the Bases. The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This ITS still retains the requirement that the core reactivity balance comparison be within +/- 1% k/k. The details of how this comparison is calculated do not need to appear in the Specification in order for the requirement to apply. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the CTS. LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 4.1.1.1.2 is applicable in MODES 1, 2, 3, and 4. ITS 3.1.2 is applicable in MODES 1 and 2. This changes the CTS DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 3 of 5 by reducing the applicable MODES in which the core reactivity requirement must be met. The purpose of CTS Surveillance 4.1.1.1.2 is to verify the core design by comparing the actual and predicted core reactivity. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analysis and licensing basis. The core reactivity balance can only be determined when the reactor is critical (MODES 1 and 2). Additionally, after performing the Surveillance once after each refueling and after 60 EFPD, the Surveillance Frequency is once per 31 EFPD, which continues to accrue when the reactor is critical. Therefore, reducing the applicable MODES from MODES 1, 2, 3, and 4 to MODES 1 and 2 does not result in a reduction of the verification of this important measure of core design accuracy. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L02 (Category 4 - Relaxation of Required Action) CTS 3.1.1.1 does not contain ACTIONS to follow if the core reactivity balance Surveillance is not met. If the core reactivity balance Surveillance is not met, CTS LCO 3.0.3 would be entered. CTS LCO 3.0.3 requires the plant to be in MODE 3 within 7 hours, MODE 4 within 13 hours, and MODE 5 within 37 hours. ITS 3.1.2 contains ACTIONS to follow if the core reactivity LCO is not met. If the LCO is not met, 7 days are provided to re-evaluate the core design and safety analysis, to determine that the reactor core is acceptable for continued operation, and to establish appropriate operating restrictions and SRs. If these actions are not completed within the 7 days, the plant must be placed in MODE 3 within 6 hours. This changes the CTS by providing 7 days to evaluate and provide compensatory measures for not meeting the core reactivity balance requirement and then requiring entry into MODE 3 instead of requiring an immediate shutdown and entry into MODE 5. The purpose of CTS 4.1.1.1.2 is to verify the accuracy of the core design by comparing the predicted and actual core reactivity throughout core life. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. Should the core reactivity balance requirement not be met, time is required to determine the cause of the disagreement and what adjustments may be needed to the operating conditions of the core. The startup physics testing program is used to verify most of the critical core design parameters, such as control rods worth, boron worth, and moderator temperature coefficient. In addition, there is considerable conservatism in the application of these values in the accident analyses. Therefore, allowing a time to evaluate the difference and make any adjustments to the operational controls is acceptable. The 7 day Completion time is reasonable considering the complexity of the evaluations and the time to meet administrative requirements, such as 10 CFR 50.59 safety evaluation DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 4 of 5 preparation and approval. If it cannot be determined within 7 days that the core is acceptable for continued operation, the unit must be shutdown. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L03 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.1.1.1.2 requires comparison of the actual and predicted core reactivity balance at least once per 31 Effective Full Power Days (EFPD) and specifically requires consideration of at least those factors stated in Specification 4.1.1.1.1.e. CTS 4.1.1.1.2 also requires the predicted reactivity values to be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. CTS 4.1.1.1.1.e requires the determination of SDM by considering the reactor coolant system boron concentration, control rod position, reactor coolant system average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, and samarium concentration in MODE 3 or 4. ITS SR 3.1.2.1 requires verifying the measured core reactivity is within +/- 1 % k/k of the predicted core reactivity values once prior to entering MODE 1 after each refueling and every 31 EFPD thereafter after 60 EFPD. This changes the CTS by not requiring the periodic, at-power core reactivity comparison until core burnup reaches 60 EFPD. Additionally, it allows the initial verification to be performed in MODE 2. The purpose of CTS 4.1.1.1.2 is to verify the agreement between the actual and predicted core reactivity. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure it provides an acceptable level of equipment reliability. The CTS and ITS require the predicted core reactivity values to be normalized to the actual values prior to exceeding 60 EFPD of core burnup. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after fuel loading, is acceptable, based on the slow rate of core reactivity changes resulting from fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly. In addition, CTS 4.1.1.1.1.e Frequency has been changed to ensure core reactivity is within limits prior to entering MODE 1 after each refueling. This change has been designated as less restrictive because Surveillances will be performed less frequently and in different MODES of operation under the ITS than under the CTS. L04 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 4.1.1.1.2 requires, in part, that the predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days (EFPD) after each fuel loading. ITS SR 3.1.2.1 contains an SR Note that states the adjustment "may" be performed prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. This changes the CTS by stating that the normalization may be performed prior to 60 EFPD after each fuel loading. The purpose of adjusting the predicted reactivity values to the core conditions is to allow benchmarking of the design calculations. Making this adjustment 60 EFPD of operation allows sufficient time for the core conditions to reach DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 5 of 5 steady state. This change is acceptable because the expectation is to perform the adjusting of the predicted reactivity values to the core conditions. ITS SR 3.1.2.1 still allows the adjustment to take place prior to the 60 EFPD after each fuel loading. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Core Reactivity 3.1.2 Westinghouse STS 3.1.2-1 Rev. 4.0 CTS 1SEQUOYAH UNIT 1 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Core Reactivity

LCO 3.1.2 The measured core reactivity shall be within +/- 1% k/k of predicted values. APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Measured core reactivity not within limit. A.1 Re-evaluate core design and safety analysis, and determine that the reactor core is acceptable for continued operation. AND A.2 Establish appropriate operating restrictions and SRs. 7 days

7 days B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. 6 hours DOC A02 Applicability DOC L02 DOC L02 Core Reactivity 3.1.2 Westinghouse STS 3.1.2-2 Rev. 4.0 CTS 1SEQUOYAH UNIT 1 Amendment XXX SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 ---------------------------NOTE---------------------------------- The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading. --------------------------------------------------------------------- Verify measured core reactivity is within +/- 1% k/k of predicted values.

Once prior to entering MODE 1 after each refueling AND

NOTE--------

Only required after 60 EFPD ------------------------ [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4.1.1.1.1.e, 4.1.1.1.2 22 Core Reactivity 3.1.2 Westinghouse STS 3.1.2-1 Rev. 4.0 CTS 1SEQUOYAH UNIT 2 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Core Reactivity

LCO 3.1.2 The measured core reactivity shall be within +/- 1% k/k of predicted values. APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Measured core reactivity not within limit. A.1 Re-evaluate core design and safety analysis, and determine that the reactor core is acceptable for continued operation. AND A.2 Establish appropriate operating restrictions and SRs. 7 days

7 days B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. 6 hours DOC A02 Applicability DOC L02 DOC L02 Core Reactivity 3.1.2 Westinghouse STS 3.1.2-2 Rev. 4.0 CTS 1SEQUOYAH UNIT 2 Amendment XXX SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 ---------------------------NOTE---------------------------------- The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading. --------------------------------------------------------------------- Verify measured core reactivity is within +/- 1% k/k of predicted values.

Once prior to entering MODE 1 after each refueling AND

NOTE--------

Only required after 60 EFPD ------------------------ [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4.1.1.1.1.e, 4.1.1.1.2 22 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 2. ISTS SR 3.1.2.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Core Reactivity B 3.1.2 WOG B 3.1.2-1 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.2 Core Reactivity

BASES BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable, such that subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity. Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the Reactor Coolant System (RCS) versus cycle burnup. Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations and that the calculational models used to generate the safety analysis are adequate.

In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration. 5specific Core Reactivity B 3.1.2 WOG B 3.1.2-2 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

BACKGROUND (continued) When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated. APPLICABLE The acceptance criteria for core reactivity are that the reactivity balance SAFETY limit ensures plant operation is maintained within the assumptions of ANALYSES the safety analyses. Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity. Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion. The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred. 1life (BOL) 1BOL 1BOL Core Reactivity B 3.1.2 WOG B 3.1.2-3 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

APPLICABLE SAFETY ANALYSES (continued) The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle. Core reactivity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of +/- 1% k/k has been established based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated. When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely. APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing. In MODE 6, fuel loading results in a continually changing core reactivity. Boron concentration requirements (LCO 3.9.1, "Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). 51BOL Core Reactivity B 3.1.2 WOG B 3.1.2-4 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined. The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.

B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

Core Reactivity B 3.1.2 WOG B 3.1.2-5 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. [ The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2. FSAR, Chapter [15]. U 2321426, if required, may 1BOL Core Reactivity B 3.1.2 WOG B 3.1.2-1 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.2 Core Reactivity

BASES BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable, such that subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity. Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the Reactor Coolant System (RCS) versus cycle burnup. Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations and that the calculational models used to generate the safety analysis are adequate.

In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration. 5specific Core Reactivity B 3.1.2 WOG B 3.1.2-2 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

BACKGROUND (continued) When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated. APPLICABLE The acceptance criteria for core reactivity are that the reactivity balance SAFETY limit ensures plant operation is maintained within the assumptions of ANALYSES the safety analyses. Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity. Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion. The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred. 1life (BOL) 1BOL 1BOL Core Reactivity B 3.1.2 WOG B 3.1.2-3 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

APPLICABLE SAFETY ANALYSES (continued) The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle. Core reactivity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of +/- 1% k/k has been established based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated. When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely. APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing. In MODE 6, fuel loading results in a continually changing core reactivity. Boron concentration requirements (LCO 3.9.1, "Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). 51BOL Core Reactivity B 3.1.2 WOG B 3.1.2-4 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined. The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.

B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

Core Reactivity B 3.1.2 WOG B 3.1.2-5 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. [ The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2. FSAR, Chapter [15]. U 2321426, if required, may 1BOL JUSTIFICATION FOR DEVIATIONS ITS 3.1.2 BASES, CORE REACTIVITY Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. ISTS SR 3.1.2.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. 3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 4. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 5. Editorial changes made for enhanced clarity/consistency. 6. Changes are made to be consistent with changes made to the Specification.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.2, CORE REACTIVITY Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 3 ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.3 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the COLR. The maximum upper limit shall be less than 0 delta k/k/°F. APPLICABILITY: Beginning of cycle life (BOL) limit - MODES 1 and 2* only# End of life cycle (EOL) limit - MODES 1, 2 and 3 only# ACTION:

a. With the MTC more positive than the BOL limit specified in the COLR operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6. 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition. b. With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours.

________________________ *With Keff greater than or equal to 1.0 #See Special Test Exception 3.10.3

May 24, 2002 SEQUOYAH - UNIT 1 3/4 1-4 Amendment No. 36, 155, 276 LCO 3.1.3 Applicability Applicability ACTION A, ACTION B ACTION C ACTION A ACTION B A02A03MODE 2 with keff <1.0A04L01A02Page 1 of 4 A01ITS ITS 3.1.3 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 PPM surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates that MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured and compared to the EOL MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuel cycle.

October 23, 1991 SEQUOYAH - UNIT 1 3/4 1-5 Amendment No. 155 SR 3.1.3.1, SR 3.1.3.2 Add proposed SR 3.1.3.2 Note 3L02Page 2 of 4 SR 3.1.3.1 SR 3.1.3.2, SR 3.1.3.2 Notes 1 and 2 A01ITS ITS 3.1.3 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the COLR. The maximum upper limit shall be less than 0 delta k/k/°F. APPLICABILITY: Beginning of Cycle life (BOL) Limit - Modes 1 and 2* only# End of Cycle Life (EOL) Limit - Modes 1, 2, and 3 only# ACTION: a. With the MTC more positive than the BOL limit specified in the COLR operation in Modes 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the EOL limit specified in the COLR be in HOT SHUTDOWN within 12 hours.

With keff greater than or equal to 1.0 # See Special Test Exception 3.10.3

May 24, 2002 SEQUOYAH - UNIT 2 3/4 1-4 Amendment Nos. 28, 146, 267 LCO 3.1.3 Applicability Applicability A02A03A04L01A02Page 3 of 4 ACTION A, ACTION B ACTION A ACTION B ACTION C MODE 2 with keff <1.0 A01ITS ITS 3.1.3 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 PPM surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than 300 PPM surveillance limit specified in the COLR, the MTC shall be remeasured and compared to the EOL MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuel cycle.

March 30, 1992 SEQUOYAH - UNIT 2 3/4 1-5 Amendment No. 146 SR 3.1.3.1, SR 3.1.3.2 Add proposed SR 3.1.3.2 Note 3L02SR 3.1.3.1 Page 4 of 4 SR 3.1.3.2, SR 3.1.3.2 Notes 1 and 2 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 The Applicability of CTS 3.1.1.3 is modified by footnote # stating "See Special Test Exception 3.10.3." ITS 3.1.3 Applicability does not contain the footnote or a reference to the Special Test Exception. This changes the CTS by not including footnote # in the ITS. The purpose of the footnote reference is to alert the user that a Special Test Exception exists that may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS. A03 CTS 3.1.1.3 ACTION a.1 states that if the MTC is more positive than the BOL limit, control rod withdrawal limits must be imposed within 24 hours or the unit must be in HOT STANDBY within the next 6 hours. ITS 3.1.3 ACTION A states that with the MTC not within the BOL limit, establish administrative control rod withdrawal limits within 24 hours or ACTION B requires the unit to be in MODE 2 with keff < 1.0 within the next 6 hours. This changes the CTS by requiring the unit to be in MODE 2 with keff < 1.0 instead of HOT STANDBY (i.e., MODE 3). This change is acceptable because the requirements have not changed. In accordance with CTS LCO 3.0.1, ACTIONS are only required to be followed while in the MODE of Applicability. The CTS BOL MTC limit is only applicable in MODE 1 and MODE 2 with keff 1.0. Therefore, under the CTS, the unit does not have to enter MODE 3 because the applicability of the ACTION ends when in MODE 2 with keff < 1.0. As a result, there is no difference between the CTS and ITS requirements. This change is designated as administrative because it does not result in a technical change to the CTS. A04 CTS 3.1.1.3 ACTION a.1 states that if the MTC is more positive than the BOL limit, then control rod withdrawal limits must be established. It also states that these withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6. ITS 3.1.3 does not contain this statement. This changes the CTS by not including the statement that the withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6. This change is acceptable because the requirements have not changed. The CTS reference to Specification 3.1.3.6 is an "information only" statement that neither adds, eliminates, or modifies requirements. The ITS convention is to not DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and Unit 2 Page 2 of 3 include these types of statements. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES None

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.1.3 ACTION a.2 states that if the measured MTC is more positive than the BOL limit, then the control rod withdrawal limits established in ACTION a.1 must be maintained until subsequent calculation verifies that the MTC has been restored to within limits for all the rods withdrawn condition. ITS 3.1.3 does not contain a requirement that the control rod withdrawal limits must be maintained until MTC is confirmed to be within its limit by measurement. However, ITS LCO 3.0.2 states that the Required Actions shall be followed until the LCO is met or no longer applicable. The ITS 3.1.3 Bases state that physics calculations may be used to determine the time in cycle life at which the calculated MTC will meet the LCO requirement, and at this point in core life the condition may be exited and the control rod withdrawal limits removed. This changes the CTS by eliminating the requirement to verify the MTC to be within its limit before removing the control rod withdrawal limits. The purpose of CTS 3.1.1.3 ACTION a.2 is to ensure that the additional operational restrictions required to maintain the MTC within the assumptions in the safety analyses are maintained until the MTC value without the restrictions is within the LCO limits. This change is acceptable because the deleted Action is not necessary to verify that the values used to meet the LCO are consistent with the safety analyses. Thus, appropriate values continue to be tested in a manner and at a Frequency necessary to give confidence that the assumptions in the safety analyses are protected. The measurement of the MTC, boron endpoint, and control rod worth prior to entering MODE 1 is sufficient to verify, the nuclear design so that it can be accurately predicted when the all rods out, full power equilibrium MTC is within the LCO limit. Performing another measurement of beginning of cycle MTC to confirm this prediction is not necessary to give confidence that MTC is within its limit. This change is designated as less restrictive because Actions that are required in the CTS will not be required in the ITS. DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and Unit 2 Page 3 of 3 L02 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.1.1.3.b requires MTC to be determined within limits. MTC shall be measured at any THERMAL POWER within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. The measured value shall be compared to the 300 ppm Surveillance limit specified in the COLR. In the event this comparison indicates that the MTC is more negative than 300 PPM surveillance limit specified in the COLR, MTC shall be remeasured and compared to the EOL MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuel cycle. ITS SR 3.1.3.2 requires verifying MTC is within the EOL limit once each cycle. Additionally, ITS SR 3.1.3.2 is modified by three notes. The first Note states that ITS SR 3.1.3.2 is not required to be performed until 7 EFPD after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm. The second Note states that if the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, then ITS SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle. The third Note states that ITS SR 3.1.3.2 does not need to be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR. This changes the CTS by eliminating the requirement to verify that MTC is met at least once per 14 EFPD if the measured MTC at the equivalent of equilibrium RTP-ARO boron concentration of 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR. The purpose of CTS 4.1.1.3.b is to periodically verify that the MTC EOL limit is within limit if the 300 ppm Surveillance limit in the COLR is not met. This change is acceptable because the Surveillance Frequency has been evaluated to ensure it will provide an acceptable level of assurance that the MTC EOL limit is not exceeded. This will help ensure that the MTC EOL limit is not exceeded for the remainder of the cycle. The new 60 ppm Surveillance limit will be incorporated into the COLR. This new limit is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance limit, then the MTC EOL limit will not be exceeded because the gradual manner in which MTC changes with core burnup. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) MTC 3.1.3 Westinghouse STS 3.1.3-1 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC)

LCO 3.1.3 The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be [ [ ] k/k°F at hot zero power] [that specified in Figure 3.1.3-1]. APPLICABILITY: MODE 1 and MODE 2 with keff 1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. MTC not within upper limit. A.1 Establish administrative withdrawal limits for control banks to maintain MTC within limit. 24 hours B. Required Action and associated Completion Time of Condition A not met. B.1 Be in MODE 2 with keff < 1.0. 6 hours C. MTC not within lower limit. C.1 Be in MODE 4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Verify MTC is within upper limit. Prior to entering MODE 1 after each refueling 1< 03.1.1.3 Applicability ACTION a.1 ACTION a.1 ACTION b 4.1.1.3.a beginning of cycle life (BOL)end of cycle life (EOL)BOL 2222EOL BOL MTC 3.1.3 Westinghouse STS 3.1.3-2 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.3.2 ---------------------------NOTES--------------------------------

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle. 3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR. ---------------------------------------------------------------------

Verify MTC is within lower limit.

Once each cycle 4.1.1.3.b EOL2 MTC 3.1.3 Westinghouse STS 3.1.3-3 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1

Figure 3.1.3 - 1 (page 1 of 1) Moderator Temperature Coefficient Vs. Rated Thermal Power 3 MTC 3.1.3 Westinghouse STS 3.1.3-1 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC)

LCO 3.1.3 The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be [ [ ] k/k°F at hot zero power] [that specified in Figure 3.1.3-1]. APPLICABILITY: MODE 1 and MODE 2 with keff 1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. MTC not within upper limit. A.1 Establish administrative withdrawal limits for control banks to maintain MTC within limit. 24 hours B. Required Action and associated Completion Time of Condition A not met. B.1 Be in MODE 2 with keff < 1.0. 6 hours C. MTC not within lower limit. C.1 Be in MODE 4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Verify MTC is within upper limit. Prior to entering MODE 1 after each refueling 1< 03.1.1.3 Applicability ACTION a.1 ACTION a.1 ACTION b 4.1.1.3.a beginning of cycle life (BOL)end of cycle life (EOL)BOL 2222EOL BOL MTC 3.1.3 Westinghouse STS 3.1.3-2 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.3.2 ---------------------------NOTES--------------------------------

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle. 3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR. ---------------------------------------------------------------------

Verify MTC is within lower limit.

Once each cycle 4.1.1.3.b EOL2 MTC 3.1.3 Westinghouse STS 3.1.3-3 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2

Figure 3.1.3 - 1 (page 1 of 1) Moderator Temperature Coefficient Vs. Rated Thermal Power 3 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS 3.1.3 contains Figure 3.1.3-1 for Moderator Temperature Coefficient Vs Rated Thermal Power. This figure is not maintained in ITS 3.1.3. ITS 3.1.3 lists the maximum upper limit value in the LCO. Therefore, ISTS Figure 3.1.3-1 is not required and has been deleted.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) MTC B 3.1.3 Westinghouse STS B 3.1.3-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Both initial and reload cores are designed so that the beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit. The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses. 1Ulife (BOL) 1life (EOL) BOL 1EOL 1 MTC B 3.1.3 Westinghouse STS B 3.1.3-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

BACKGROUND (continued) If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity. The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup. APPLICABLE The acceptance criteria for the specified MTC are: SAFETY ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2) and

b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3). The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 4) or RTP, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature. 1U MTC B 3.1.3 Westinghouse STS B 3.1.3-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

APPLICABLE SAFETY ANALYSES (continued) In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2). MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.

MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration. LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation. Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and more positive than a given lower bound. The MTC is most positive at BOC; this upper bound must not be exceeded. This maximum upper limit occurs at BOC, all rods out (ARO), hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

During operation, therefore, the conditions of the LCO can only be ensured through measurement. The Surveillance checks at BOC and EOC on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met. 4maintainedBOL or EOL 1BOL and EOL 1EOL EOL 1BOL EOL BOL BOL EOL 1 MTC B 3.1.3 Westinghouse STS B 3.1.3-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

LCO (continued) The LCO establishes a maximum positive value that cannot be exceeded. The BOC positive limit and the EOC negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule. APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above. In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled control rod assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES. ACTIONS A.1 If the BOC MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits. The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits. As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect. 35BOL EOL 51BOL EOL 1BOL MTC B 3.1.3 Westinghouse STS B 3.1.3-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

ACTIONS (continued) B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours, the unit must be brought to MODE 2 with keff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. C.1 Exceeding the EOC MTC limit means that the safety analysis assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours.

The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

The BOC MTC value for ARO will be inferred from isothermal temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks. 1BOL 1EOL EOL EOL 1BOL1BOL BOL 3at least MTC B 3.1.3 WOG B 3.1.3-6 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.

SR 3.1.3.2 is modified by three Notes that include the following requirements:

a. The SR is not required to be performed until 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm. b. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC. Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit.

c. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance limit, the EOC limit will not be exceeded because of the gradual manner in which MTC changes with core burnup. REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.

2. FSAR, Chapter [15].

3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

4. FSAR, Chapter [15]. 1212U U .2.11EOL 1EOL EOL 1EOL EOL 1EOL EOL BAW 10169P-A, "B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989 1 MTC B 3.1.3 Westinghouse STS B 3.1.3-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Both initial and reload cores are designed so that the beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit. The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses. 1Ulife (BOL) 1life (EOL) BOL 1EOL 1 MTC B 3.1.3 Westinghouse STS B 3.1.3-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

BACKGROUND (continued) If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity. The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup. APPLICABLE The acceptance criteria for the specified MTC are: SAFETY ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2) and

b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3). The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 4) or RTP, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature. 1U MTC B 3.1.3 Westinghouse STS B 3.1.3-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

APPLICABLE SAFETY ANALYSES (continued) In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2). MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.

MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration. LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation. Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and more positive than a given lower bound. The MTC is most positive at BOC; this upper bound must not be exceeded. This maximum upper limit occurs at BOC, all rods out (ARO), hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

During operation, therefore, the conditions of the LCO can only be ensured through measurement. The Surveillance checks at BOC and EOC on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met. 4maintainedBOL or EOL 1BOL and EOL 1EOL EOL 1BOL EOL BOL BOL EOL 1 MTC B 3.1.3 Westinghouse STS B 3.1.3-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

LCO (continued) The LCO establishes a maximum positive value that cannot be exceeded. The BOC positive limit and the EOC negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule. APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above. In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled control rod assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES. ACTIONS A.1 If the BOC MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits. The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits. As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect. 35BOL EOL 51BOL EOL 1BOL MTC B 3.1.3 Westinghouse STS B 3.1.3-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

ACTIONS (continued) B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours, the unit must be brought to MODE 2 with keff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. C.1 Exceeding the EOC MTC limit means that the safety analysis assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours.

The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

The BOC MTC value for ARO will be inferred from isothermal temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks. 1BOL 1EOL EOL EOL 1BOL1BOL BOL 3at least MTC B 3.1.3 WOG B 3.1.3-6 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.

SR 3.1.3.2 is modified by three Notes that include the following requirements:

a. The SR is not required to be performed until 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm. b. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC. Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit.

c. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance limit, the EOC limit will not be exceeded because of the gradual manner in which MTC changes with core burnup. REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.

2. FSAR, Chapter [15].

3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

4. FSAR, Chapter [15]. 1212U U .2.11EOL 1EOL EOL 1EOL EOL 1EOL EOL BAW 10169P-A, "B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989 1 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3 BASES, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. Editorial changes made for enhanced clarity/consistency. 4. Changes are made to be consistent with the Specification. 5. Changes are made to be consistent with changes made to the Specification.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 4 ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1* and 2* ACTION: a. With one or more full length rods untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours. b. With more than one full length rod misaligned from the group step counter demand position by more than +/- 12 steps (indicated position), be in HOT STANDBY within 6 hours. c. With one full length rod misaligned from its group step counter demand height by more than +/- 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1. The rod is restored within the above alignment requirements, or 2. The remainder of the rods in the group with the misaligned rod are aligned to within +/- 12 steps of the misaligned rod while maintaining the rod sequence and insertion limit of specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

____________________ *See Special Test Exceptions 3.10.2 and 3.10.3. November 21, 1995 SEQUOYAH - UNIT 1 3/4 1-14 Amendment No. 114, 155, 215 Add proposed Required Action A.1.2L01LCO 3.1.4 Page 1 of 12 Applicability ACTION A ACTION D Add proposed Required Action B.2.1.2 L01A03A02L02A02ACTION B Add proposed Required Action D.1.1 and D.1.2 M01Alignment Limits A01Rod ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS ACTION: (Continued) a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. c) A power distribution map is obtained from the movable incore detectors and FQ(Z) and FNH are verified to be within their limits within 72 hours. d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be trippable by verifying rod freedom of movement by movement of 10 steps in either direction at least once per 92 days.

November 21, 1995 SEQUOYAH - UNIT 1 3/4 1-15 Amendment No. 215 LA03L04Page 2 of 12 ACTION B Add proposed ACTION CM02L03twoL05SR 3.1.4.1 SR 3.1.4.2 LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program ITS ITS 3.1.4 A01 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH ROD

Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident) Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

SEQUOYAH - UNIT 1 3/4 1-16 LA03Page 3 of 12 ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position# shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. Tavg greater than or equal to 541°F, and b. All reactor coolant pumps operating. APPLICABILITY: MODES 1 and 2 ACTION: a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 71% of RATED THERMAL POWER SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months. _____________ #Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of 222 and 231 steps withdrawn, inclusive.

May 08, 1990 SEQUOYAH - UNIT 1 3/4 1-19 Amendment No. 108, 138 Page 4 of 12 500L06A04L07L08LA02SR 3.1.4.3 SR 3.1.4.3 Add proposed ACTION A M04M03Applicability LA02 ITS ITS 3.1.4 A013/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg Greater Than 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

____________________ *See Special Test Exception 3.10.1 November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-1 Amendment No. 172 Page 5 of 12 See ITS 3.1.1 See ITS 3.1.1 See ITS Chapter 1.0See ITS 3.1.1 L09 ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-3 Amendment No. 12, 172 Page 6 of 12 See ITS 3.1.1 See ITS 3.1.1 See ITS Chapter 1.0L09 ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: Modes 1* and 2*. ACTION:

a. With one or more full length rods untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours. b. With more than one full length rod misaligned from the group step counter demand position by more than +/- 12 steps (indicated position), be in HOT STANDBY within 6 hours. c. With one full length rod misaligned from its group step counter demand height by more than +/- 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either: 1. The rod is restored within the above alignment requirements, or 2. The remainder of the rods in the group with the misaligned rod are aligned to within +/- 12 steps of the misaligned rod while maintaining the rod sequence and insertion limit of specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.

See Special Test Exceptions 3.10.2 and 3.10.3.

November 21, 1995 SEQUOYAH - UNIT 2 3/4 1-14 Amendment Nos. 104, 146, 205 Add proposed Required Action A.1.2L01LCO 3.1.4 Page 7 of 12 Applicability Add proposed Required Action B.2.1.2L01A03A02L02A02Add proposed Required Action D.1.1 and D.1.2 M01ACTION A ACTION D ACTION B LA03Alignment Limits A01Rod ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS ACTION: (Continued)

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. c) A power distribution map is obtained from the movable incore detectors FQ (Z) and NHFare verified to be within their limits within 72 hours. d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be trippable by verifying rod freedom of movement by movement of 10 steps in either direction at least once per 92 days.

November 21, 1995 SEQUOYAH - UNIT 2 3/4 1-15 Amendment No. 205 L04ACTION B Add proposed ACTION CM02L03twoL05SR 3.1.4.1 SR 3.1.4.2 LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program Page 8 of 12 ITS ITS 3.1.4 A01TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident) Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

SEQUOYAH - UNIT 2 3/4 1-16 LA03Page 9 of 12 ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position# shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: a. Tavg greater than or equal to 541°F, and b. All reactor coolant pumps operating. APPLICABILITY: Modes 1 and 2. ACTION: a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 71% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head, b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

    # Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of >222 and <231 steps withdrawn, inclusive.

October 4, 1995 SEQUOYAH - UNIT 2 3/4 1-19 Amendment Nos. 20, 98, 130, 203 500L06A04L07L08LA02SR 3.1.4.3 SR 3.1.4.3 Add proposed ACTION A M04M03Page 10 of 12Applicability LA02 ITS ITS 3.1.4 A013/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2, with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-1 Amendment No. 163 See ITS 3.1.1 See ITS 3.1.1 See ITS Chapter 1.0See ITS 3.1.1 L09Page 11 of 12 ITS ITS 3.1.4 A01REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg Less Than or Equal to 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration,

2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-3 Amendment No. 163

See ITS 3.1.1 See ITS 3.1.1 See ITS Chapter 1.0L09Page 12 of 12 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 10 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.1.3.1 Applicability is modified by Footnote

which states "See Special Test Exceptions 3.10.2 and 3.10.3." ITS 3.1.4 Applicability does not contain this Note. This changes the CTS by not including Footnote *.

The purpose of Footnote

is to alert the Technical Specification user that a Special Test Exception exists that may modify the Applicability of this Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.1.3.1 ACTION c.2 states that with one full length rod misaligned from its group step counter demand height by more than +/- 12 steps (indicated position), POWER OPERATION may continue provided that within one hour, the remainder of the rods in the group with the misaligned rod are aligned to within +/- 12 steps of the misaligned rod while maintaining the rod sequence and insertion limit of specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation. ITS 3.1.4 does not contain a Required Action stating that the remainder of the rods in the group must be aligned with the misaligned rod. This changes the CTS by not including a specific Required Action stating that the remainder of the rods in the group must be aligned with the misaligned rod. This change is acceptable because the technical requirements have not changed. The moving of the remaining rods to within the LCO limit of the misaligned rod, while complying with all of the other rod position requirements, is simply restoring compliance with the LCO. Restoration of compliance with the LCO is always an available Required Action and it is the convention of the ITS to not state such "restore" options explicitly unless it is the only action or is required for clarity. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 3.1.3.4 ACTION a states with the drop time of any full length rod determined to exceed the above limit restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. ITS 3.1.4 does not have a similar requirement. This changes the CTS by not explicitly requiring, in the ITS 3.1.4 ACTIONS, restoration of the rod drop time prior to proceeding to MODE 1 or 2. DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 2 of 10 CTS 4.0.4 and ITS SR 3.0.4 require verification that Surveillances are met prior to entering the MODE in which they apply. CTS 4.0.4 and ITS SR 3.0.4 also prohibit entering a MODE or condition with the Surveillance not met and while relying on actions. Therefore, since the Applicability of CTS 3.1.3.4 is MODES 1 and 2, the action prohibiting entry into MODES 1 and 2 with the rod drop time requirements not met is redundant to CTS 4.0.4 and ITS 3.0.4. This change is acceptable because the technical requirements have not changed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 3.1.3.1 ACTION b states "With more than one full length rod misaligned from the group step counter demand position by more than +/- 12 steps (indicated position), be in HOT STANDBY within 6 hours." ITS 3.1.4 ACTION D adds additional requirements (ITS 3.1.4 Required Actions D.1.1 and D.1.2) to verify SHUTDOWN MARGIN is within the limits within 1 hour or to initiate boration to restore the required SHUTDOWN MARGIN to within limits. This changes the CTS by adding two additional Required Actions. The purpose of CTS 3.1.3.1 ACTION a is to place the unit in a MODE in which the equipment is not required. More than one control rod misaligned from its group average has the potential to reduce the SHUTDOWN MARGIN. Therefore, the SHUTDOWN MARGIN must be evaluated. ITS 3.1.4 adds Required Actions to allow verification that the SHUTDOWN MARGIN is within the limit or to borate to restore the SHUTDOWN MARGIN to within limits. These new Required Actions must be accomplished within 1 hour. The one hour allows the operator adequate time to determine the SHUTDOWN MARGIN. Restoration of the required SHUTDOWN MARGIN, if necessary, requires increasing the RCS boron concentration to provide negative reactivity. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete this action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SHUTDOWN MARGIN is restored. This change is acceptable because it is consistent with the assumptions of the safety analyses to be within the SHUTDOWN MARGIN limit. This change has been designated as more restrictive because it adds explicit actions to verify SHUTDOWN MARGIN or to restore SHUTDOWN MARGIN within limits. M02 CTS 3.1.3.1 ACTION c requires that with one full length rod misaligned, POWER OPERATION may continue provided certain actions are completed within one hour. If those actions are not complete, CTS 3.0.3 is required to be entered since no further actions are specified. CTS 3.0.3 allows 1 hour to initiate action and 6 additional hours for the unit to be placed in MODE 3. ITS 3.1.4 ACTION C states that if the Required Action and associated Completion Time of Condition B is not met, the unit must be in MODE 3 within 6 hours. This changes the CTS by providing a specific default condition instead of requiring entry into CTS 3.0.3, and thereby reduces the time to reach MODE 3 following discovery of a misaligned rod if Required Actions are not met from 7 hours to 6 hours. DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 3 of 10 The purpose of requiring a shutdown when a rod misalignment cannot be corrected is to bring the unit to a subcritical condition prior to the buildup of an undesirable reactor core power distribution. This change is acceptable because the proposed default condition will require the plant to be in a condition where the rod group alignment limits are no longer applicable. The proposed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power in an orderly manner and without challenging unit systems. This change is designated as more restrictive since the 1 hour specified in CTS 3.0.3 no longer applies. M03 CTS 3.1.3.4 ACTION b provides an allowance for operation to proceed with THERMAL POWER restricted to less than or equal to 71% of RATED THERMAL POWER, with rod drop times within limits but determined with 3 reactor coolant pumps operating. ITS 3.1.4 does not contain a similar allowance. This changes the CTS by not allowing continued operation at reduce power when the rod drop times are determined with only 3 reactor coolant pumps operating. The purpose of CTS 3.1.3.4 is to ensure the rods insert within the rod drop criteria. This change is acceptable because ITS SR 3.1.4.3 requires verification of the rod drop times be performed with all of the RCPs operating and the average moderator temperature is 500°F. Therefore, ITS 3.1.4 will not allow the rod drop times to be determined with only 3 reactor coolant pumps operating. This change is designated as more restrictive because an allowance is being removed from the CTS. M04 CTS 3.1.3.4 ACTION a requires that with the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. CTS 3.1.3.4 ACTION b requires that with the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 71% of RATED THERMAL POWER. However, no specific actions are stated in CTS 3.1.3.4 when the unit is in MODES 1 and 2 when the drop time is discovered to not be within limits. Therefore, CTS 3.0.3 entry would be required. CTS 3.0.3 allows one hour to prepare for a shutdown and requires the unit to be in HOT STANDBY (MODE 3) within 7 hours. ITS 3.1.4 ACTION A applies with one or more rods inoperable. ITS 3.1.4 ACTION A requires verification that the SDM is within the limits specified in the COLR or initiate boration to restore the SDM to within limit within one hour, and to be in MODE 3 within 6 hours. This changes the CTS by adding new requirements associated with SDM and changing the requirement to be outside of the MODE of Applicability from 7 hours to 6 hours. The purpose of requiring a shutdown when a drop time of any full length rod is not met is to bring the unit to a subcritical condition. With one or more slow control rod(s) there is a potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution in the reactor core, the low probability of an DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 4 of 10 accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored. In addition, the new time to reach MODE 3 is consistent with the time provided in other specifications. This change is acceptable because it is consistent with the requirements of the assumptions of the safety analyses to be within the SDM limit. The change has been designated as more restrictive because it adds explicit actions to verify SDM or to restore SDM within limits and reduces the time required to be in MODE 3. RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.1.3.1.1 requires that the position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours. CTS 4.1.3.1.2 requires each full-length rod not fully inserted in the core shall be determined to be trippable by verifying rod freedom of movement by movement of 10 steps in either direction at least once per 92 days. ITS SR 3.1.4.1 and SR 3.4.1.2 require similar Surveillances and specify the periodic Frequencies as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.3.4 requires the individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 5 of 10 dashpot entry with Tavg greater than or equal to 541°F and all reactor coolant pumps operating. Additionally, it contains a footnote (footnote #) which states "Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of >222 and <231 steps withdrawn, inclusive." ITS 3.1.4 does not contain the footnote. This changes the CTS by relocating the footnote to the Bases. The removal of these details, that are related to system design, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement for performing rod drop time testing from the fully withdrawn position. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.1.3.1 ACTION c.3.a) states when a rod is misaligned, POWER OPERATION may continue if a reevaluation of each accident analysis in Table 3.1-1 is performed within 5 days. This reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. ITS 3.1.4 Required Action B.2.6 states that when one rod is misaligned, re-evaluate the safety analyses and confirm results remain valid for the duration of operation under these conditions. This changes the CTS by moving the accidents listed in Table 3.1-1 to the UFSAR. The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to re-evaluate the safety analyses and confirm results remain valid for the duration of operation under these conditions. Additionally, this change is acceptable because the removed information will be adequately controlled in the UFSAR. The UFSAR is controlled under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to procedural detail is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.3.1 ACTION a states, in part, with one or more full length rods untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour. CTS 3.1.3.1 ACTION c.3 states, in part, with one full length rod misaligned from its group step counter demand height by more than +/- 12 steps (indicated DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 6 of 10 position), the rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour. ITS 3.1.4 ACTION A and B requires, within 1 hour, to verify SHUTDOWN MARGIN is within the limits specified in the COLR or to initiate boration to restore SDM to within limits. This changes the CTS by allowing boration to restore SHUTDOWN MARGIN. The purpose of CTS 3.1.3.1 ACTION a and c.3 is to verify adequate SHUTDOWN MARGIN exists. This change is acceptable because the ITS 3.1.4 Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair the inoperable features. When a rod is inoperable or misaligned, boration may be required to reestablish compliance with the SHUTDOWN MARGIN requirements. Providing a short period of time to reestablish the SHUTDOWN MARGIN requirement instead of entering ITS LCO 3.0.3 is justified because of the existing conservatisms in the SHUTDOWN MARGIN calculations. This change has been designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L02 (Category 4 - Relaxation of Required Action) CTS 3.1.3.1 ACTION c specifies the requirements for one full length rod misaligned from its group step counter demand height by more than the allowed rod alignment. CTS 3.1.3.1 ACTION c.3 requires the affected rod to be declared inoperable. ITS 3.1.4 ACTION B specifies requirements for one rod not within alignment limits and does not require that the rod be declared inoperable. This changes the CTS by deleting the requirement to declare a misaligned rod inoperable. The purpose of ITS 3.1.4 is to ensure that the shutdown and control rods are capable of performing their safety function of inserting into the core when required. A secondary function of the control rods is to maintain alignment so that the reactor core power distribution is consistent with the safety analyses. This change is acceptable because the LCO requirements continue to ensure that structures, systems, and components are maintained consistent with the safety analyses and licensing basis. In the ITS, rod OPERABILITY is related only to trippability, and a misaligned rod is not considered inoperable if it can be tripped. Misalignment is addressed by the ITS 3.1.4 LCO, but is separate from OPERABILITY. In both cases, trippability and misalignment, the ITS continues to provide appropriate compensatory measures. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L03 (Category 4 - Relaxation of Required Action) CTS 3.1.3.1 ACTION c.3.d) states that with one rod misaligned, reduce the THERMAL POWER level to less than 75% of the RATED THERMAL POWER within one hour. ITS 3.1.4 Required Action B.2.2 requires THERMAL POWER to be reduced to 75% of the RATED THERMAL POWER within two hours. This changes the CTS by changing the Completion Time from one hour to two hours. The purpose of CTS 3.1.3.1 ACTION c.3.d) is to reduce reactor core power to ensure that the increases in linear heat generation rate due to misalignment of a DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 7 of 10 rod does not result in exceeding the design limits. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, the capacity and capability of remaining features, and the low probability of a DBA occurring during the allowed Completion Time. The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Trip System. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS. L04 (Category 4 - Relaxation of Required Action) CTS 3.1.3.1 ACTION c.3.d) states that with one rod misaligned, reduce the high neutron flux setpoint to less than or equal to 85% of RATED THERMAL POWER within the next 4 hours. ITS 3.1.4 Required Action B.2.2 requires THERMAL POWER to be reduced to 75% RTP, but does not require the high neutron flux trip setpoint to be reduced. This changes the CTS by eliminating the Required Action to reduce the high neutron flux trip setpoint. The purpose of CTS 3.1.3.1 ACTION c.3.d) is to reduce reactor core power to ensure that the increases in linear heat generation rate due to misalignment of a rod does not result in exceeding the design limits. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, the capacity and capability of remaining features, and a low probability of a DBA occurring during the repair period. Lowering the high neutron flux trip setpoint increases the chance of an inadvertent reactor trip due to the changes being made to the Reactor Trip System without providing commensurate amount of added safety. Administrative methods of maintaining reactor power below that allowed by the Required Action are sufficient to protect the core. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L05 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.1.3.1.1 states that the position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then verifying the group positions at least once per 4 hours. ITS SR 3.1.4.1 requires verifying individual rod positions are within alignment limits in accordance with the Surveillance Frequency Control Program. This changes the CTS by eliminating the requirements to verify the individual rod position to be within alignment limits every 4 hours when the Rod Position Deviation Monitor is inoperable. See DOC LA01 for the relocation of the CTS 4.1.3.1.1 Frequency to the Surveillance Frequency Control Program. The purpose of CTS 4.1.3.1.1 is to periodically verify that the rods are within the alignment limits specified in the LCO. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of rod position verification when the Rod Position Deviation Monitor is inoperable is unnecessary, since an inoperability of the alarm does not increase the probability DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 8 of 10 that the rods are misaligned. The Rod Deviation Monitor, as described in the safety analysis is indication only and is not credited for any automatic action; however, it is there to alert the operator to a dropped rod or misaligned rod by more than 5% span. Its use is not credited in the safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L06 (Category 1 - Relaxation of LCO Requirements) CTS 3.1.3.4 requires the individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with Tavg greater than or equal to 541°F and all reactor coolant pumps operating. ITS SR 3.1.4.3 specifies the rod drop time be verified at an RCS Tavg of 500°F. This changes the CTS by lowering the required temperature at which rod drop time must be verified. The purpose of CTS 3.1.3.4 is to ensure the rods insert within the rod drop time criteria. The performance of rod drop time tests ensures that the required negative reactivity insertion (amount and rate) from a reactor trip is within the values assumed in the safety analyses. This change will allow rod drop testing to begin earlier during a startup following a refueling outage. The proposed change is acceptable because the specified rod drop time remains unchanged and the proposed 500°F test temperature is conservative compared to the CTS requirement of 541°F. Since the moderator becomes denser as the RCS temperature is decreased, a lower RCS temperature results in slower rod drops due to the density change of the water. However, the limiting rod drop time requirement of the CTS (2.7 seconds) is maintained in the ITS and must still be met. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L07 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.3.4.b requires the rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods. ITS 3.1.4 does not contain this testing requirement. This changes the CTS by not explicitly requiring post-maintenance testing on full length rods. The purpose of CTS 4.1.3.4.b is to verify OPERABILITY of the control rods following maintenance that could alter their operation. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed safety function. Any time the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post-maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1. The OPERABILITY requirements for the rod control system are described in the Bases for ITS 3.1.4. In addition, the requirements of 10 CFR 50, Appendix B, Section XI (Test Control) provide DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 9 of 10 adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. L08 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.3.4 requires drop testing of full length rods to be demonstrated through measurement prior to reactor criticality following each removal of the reactor vessel head and at least once per 18 months. ITS 3.1.4.3 requires the test to be performed prior to criticality after each removal of the reactor head. This changes the CTS by deleting the requirement to perform this test at least once per 18 months. The purpose of CTS 4.1.3.4 is to ensure the rods insert within the rod drop criteria. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its safety function. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence the equipment can perform its assumed safety function. The requirements in the CTS to perform the test following each removal of the reactor vessel head and at least once per 18 months normally coincide with one another. The head is removed once per 18 months unless there is a need to remove the head prior to the end of the cycle. This change is designated as less restrictive because a Surveillance that was required in the CTS will not be performed in the ITS. L09 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.1.1.1.a requires the SHUTDOWN MARGIN to be determined to be greater than or equal to 1.6% delta k/k within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod is inoperable. CTS 4.1.1.2.a requires the SHUTDOWN MARGIN to be determined to be greater than or equal to 1.0% delta k/k within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod is inoperable. These requirements are applicable in MODES 1, 2, 3, 4, and 5. ITS 3.1.4 Required Action A.1.1 requires the verification of SDM to be within limits within 1 hour. This verification is required in MODES 1 and 2 with one or more control rod(s) inoperable. This changes the CTS by not requiring any explicit SDM verifications for inoperable control rod(s) in MODES 3, 4, and 5 other than the normal verifications specified in ITS SR 3.1.1.1 (once every 24 hours). For MODES 1 and 2 operations, this changes the CTS by not requiring the verification of SDM on a once per 12 hour basis for one or more inoperable rod(s). The purpose of CTS 4.1.1.1.1.a and CTS 4.1.1.2.a is to provide the appropriate compensatory measures to determine SDM when control rod(s) are inoperable during operations in MODES 1, 2, 3, 4, and 5. The purpose of the ITS 3.1.4 ACTIONS are to provide the appropriate compensatory actions for inoperable control rods in MODES 1 and 2. The purpose of ITS SR 3.1.1.1 is to provide the normal Frequency for verification of SDM regardless of the status of the control rod(s). When the plant is operating in MODES 1 and 2, with one or more rod(s) inoperable, the unit must be in MODE 3 within 6 hours. After reaching MODE 3, ITS 3.1.4 no longer applies therefore it is inappropriate to specify additional DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 10 of 10 actions after the unit is outside the Applicability of the Specification. Nevertheless, SDM must still be verified in accordance with ITS SR 3.1.1.1 every 24 hours. This SDM verification must also compensate for the reactivity worth of the control rod that is not fully inserted since it is required by the definition of SDM. Therefore, ITS 3.1.4 ACTIONS provide the appropriate compensatory measures. In MODES 3 and 4, SDM will be monitored in accordance with ITS SR 3.1.1.1 every 24 hours. This change is acceptable since SDM will still be required to be monitored every 24 hours, and based on the definition of SDM the reactivity worth of any rod not capable of being fully inserted must be accounted for in the determination of SDM. Thus, SDM continues to be monitored in a manner and at a Frequency necessary to give confidence that the assumptions in the safety analyses are protected. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-1 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits

LCO 3.1.4 All shutdown and control rods shall be OPERABLE. AND Individual indicated rod positions shall be within 12 steps of their group step counter demand position.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) inoperable. A.1.1 Verify SDM to be within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Be in MODE 3. 1 hour

1 hour

6 hours B. One rod not within alignment limits. B.1 Restore rod to within alignment limits. OR B.2.1.1 Verify SDM to be within the limits specified in the COLR. OR 1 hour

1 hour 3.1.3.1 3.1.3.1 Applicability, 3.1.3.4 Applicability 3.1.3.1 ACTION a, 4.1.1.1.1, 4.1.1.2, DOC M04 3.1.3.1 ACTION c Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-2 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2.1.2 Initiate boration to restore SDM to within limit. AND B.2.2 Reduce THERMAL POWER to 75% RTP. AND B.2.3 Verify SDM is within the limits specified in the COLR. AND B.2.4 Perform SR 3.2.1.1 and SR 3.2.1.2. AND B.2.5 Perform SR 3.2.2.1.

AND B.2.6 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions. 1 hour

2 hours

Once per 12 hours

72 hours

5 days C. Required Action and associated Completion Time of Condition B not met. C.1 Be in MODE 3. 6 hours D. More than one rod not within alignment limit. D.1.1 Verify SDM is within the limits specified in the COLR. OR 1 hour 3.1.3.1 ACTION b DOC M02 3.1.3.1 ACTION c 1 Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-3 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D.1.2 Initiate boration to restore required SDM to within limit. AND D.2 Be in MODE 3.

1 hour

6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core 10 steps in either direction. [ 92 days OR In accordance with the Surveillance Frequency Control Program ] 3.1.3.1 ACTION b 4.1.3.1.1 4.1.3.1.2 2222 Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-4 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is [2.2] seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with: a. Tavg 500°F and

b. All reactor coolant pumps operating.

Prior to criticality after each removal of the reactor head 32.7 3.1.3.4, 4.1.3.4 Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-1 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits

LCO 3.1.4 All shutdown and control rods shall be OPERABLE. AND Individual indicated rod positions shall be within 12 steps of their group step counter demand position.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) inoperable. A.1.1 Verify SDM to be within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Be in MODE 3. 1 hour

1 hour

6 hours B. One rod not within alignment limits. B.1 Restore rod to within alignment limits. OR B.2.1.1 Verify SDM to be within the limits specified in the COLR. OR 1 hour

1 hour 3.1.3.1 3.1.3.1 Applicability, 3.1.3.4 Applicability 3.1.3.1 ACTION a, 4.1.1.1.1, 4.1.1.2, DOC M04 3.1.3.1 ACTION c Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-2 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2.1.2 Initiate boration to restore SDM to within limit. AND B.2.2 Reduce THERMAL POWER to 75% RTP. AND B.2.3 Verify SDM is within the limits specified in the COLR. AND B.2.4 Perform SR 3.2.1.1 and SR 3.2.1.2. AND B.2.5 Perform SR 3.2.2.1.

AND B.2.6 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions. 1 hour

2 hours

Once per 12 hours

72 hours

5 days C. Required Action and associated Completion Time of Condition B not met. C.1 Be in MODE 3. 6 hours D. More than one rod not within alignment limit. D.1.1 Verify SDM is within the limits specified in the COLR. OR 1 hour 3.1.3.1 ACTION b DOC M02 3.1.3.1 ACTION c 1 Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-3 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D.1.2 Initiate boration to restore required SDM to within limit. AND D.2 Be in MODE 3.

1 hour

6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core 10 steps in either direction. [ 92 days OR In accordance with the Surveillance Frequency Control Program ] 3.1.3.1 ACTION b 4.1.3.1.1 4.1.3.1.2 2222 Rod Group Alignment Limits 3.1.4 WOG STS 3.1.4-4 Rev. 4.0, CTS 1Amendment XXX SEQUOYAH UNIT 2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is [2.2] seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with: a. Tavg 500°F and

b. All reactor coolant pumps operating.

Prior to criticality after each removal of the reactor head 32.7 3.1.3.4, 4.1.3.4 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 2. ISTS SR 3.1.4.1 and SR 3.1.4.2 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. 3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-1 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX B 3.3 INSTRUMENTATION B 3.1.4 Rod Group Alignment Limits

BASES BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM. Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately e inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All units have four control banks and at least two shutdown banks. The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with 5/811Each unit has four1B 3.1 REACTIVITY CONTROL SYSTEMS 4 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-2 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

BACKGROUND (continued) control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System. The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI System is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy. APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY (Ref. 3). The acceptance criteria for addressing control rod inoperability ANALYSES or misalignment are that:

a. There be no violations of: 1. Specified acceptable fuel design limits or 2. Reactor Coolant System (RCS) pressure boundary integrity and b. The core remains subcritical after accident transients. Rod Position Indication5/8 111112; ;8that8anRod Position Indication 9 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-3 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

APPLICABLE SAFETY ANALYSES (continued) Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Two types of analysis are performed in regard to static rod misalignment (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps. Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5). The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved. Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor ( FQ(Z)) and the nuclear enthalpy hot channel factor (HNF) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and HNF must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and HNF to the operating limits. Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 3 4 1133A different 1INSERT 1INSERT 21D is fully1and Control +/-1 B 3.1.4 Insert Page B 3.1.4-3 INSERT 1 There are three RCCA misalignment accidents which are analyzed. They include one or more dropped RCCAs, a dropped RCCA bank, and a statically misaligned RCCA. (Ref. 4)

INSERT 2

For the dropped RCCA(s) misalignment accident, a negative reactivity insertion will result. For those dropped RCCA(s) that do not result in a reactor trip, power may be reestablished either by reactivity feedback or control bank withdrawal. Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a concern and establishing the automatic rod control mode of operation as the limiting case.

For the dropped RCCA bank misalignment accident, a reactivity insertion of greater than 500 pcm which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is then tripped. The core is not adversely affected during this period since power is decreasing rapidly. Following the reactor trip, normal shutdown procedures are followed to further cool down the plant.

11 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-4 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis. APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling. ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is adequate for determining SDM and, if necessary, for initiating emergency boration and restoring SDM.

In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth. 110% of span 14.4 linear heat rates ( )81that, except for control rod OPERABILITY testing, 1 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-5 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

ACTIONS (continued) A.2 If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. B.1 When a rod becomes misaligned, it can usually be moved and is still trippable. If the rod can be realigned within the Completion Time of 1 hour, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction. An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour gives the operator sufficient time to adjust the rod positions in an orderly manner.

B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit. In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps. Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour. The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration. (OPERABLE)481misaligned but8 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-6 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

ACTIONS (continued) B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (FQ(Z) and HNF) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible. Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 7). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met. Verifying that FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(Z) and HNF. Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis. C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0 within 6 hours, which 5 U 31144X,Y,11resulting from 1X,Y,1X,Y,FH(X,Y)FH(X,Y)FH(X,Y) Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-7 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

ACTIONS (continued) obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems. D.1.1 and D.1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0 within 6 hours.

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.4.1 REQUIREMENTS [ Verification that individual rod positions are within alignment limits at a Frequency of 12 hours provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected. OR 3 34459of 1of the control banks Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-8 Rev. 4.0 SEQUOYAH UNIT 1 1Revision XXX BASES

SURVEILLANCE REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2 with Keff 1.0, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. [ The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken. 6556541greater than or equal to in either direction Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-9 Rev. 4.0 1Revision XXXSEQUOYAH UNIT 1 BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 500°F to simulate a reactor trip under actual conditions. This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. FSAR, Chapter [15].

4. FSAR, Chapter [15].

5. FSAR, Chapter [15].

6. FSAR, Chapter [15]. 7. FSAR, Chapter [15]. U Section 15.2.3Section 15.4.2U 17171711INSERT 31installation1 B 3.1.4 Insert Page B 3.1.4-9 INSERT 3 Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of 222 and 231 steps withdrawn, inclusive. 1 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-1 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX B 3.3 INSTRUMENTATION B 3.1.4 Rod Group Alignment Limits

BASES BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM. Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately e inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All units have four control banks and at least two shutdown banks. The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with 5/811Each unit has four1B 3.1 REACTIVITY CONTROL SYSTEMS 4 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-2 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

BACKGROUND (continued) control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System. The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI System is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy. APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY (Ref. 3). The acceptance criteria for addressing control rod inoperability ANALYSES or misalignment are that:

a. There be no violations of: 1. Specified acceptable fuel design limits or 2. Reactor Coolant System (RCS) pressure boundary integrity and b. The core remains subcritical after accident transients. Rod Position Indication5/8 111112; ;8that8anRod Position Indication 9 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-3 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

APPLICABLE SAFETY ANALYSES (continued) Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Two types of analysis are performed in regard to static rod misalignment (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps. Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5). The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved. Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor ( FQ(Z)) and the nuclear enthalpy hot channel factor (HNF) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and HNF must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and HNF to the operating limits. Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 3 4 1133A different 1INSERT 1INSERT 21D is fully1and Control +/-1 B 3.1.4 Insert Page B 3.1.4-3 INSERT 1 There are three RCCA misalignment accidents which are analyzed. They include one or more dropped RCCAs, a dropped RCCA bank, and a statically misaligned RCCA. (Ref. 4)

INSERT 2

For the dropped RCCA(s) misalignment accident, a negative reactivity insertion will result. For those dropped RCCA(s) that do not result in a reactor trip, power may be reestablished either by reactivity feedback or control bank withdrawal. Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a concern and establishing the automatic rod control mode of operation as the limiting case.

For the dropped RCCA bank misalignment accident, a reactivity insertion of greater than 500 pcm which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is then tripped. The core is not adversely affected during this period since power is decreasing rapidly. Following the reactor trip, normal shutdown procedures are followed to further cool down the plant.

11 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-4 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis. APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling. ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is adequate for determining SDM and, if necessary, for initiating emergency boration and restoring SDM.

In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth. 110% of span 14.4 linear heat rates ( )81that, except for control rod OPERABILITY testing, 1 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-5 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

ACTIONS (continued) A.2 If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. B.1 When a rod becomes misaligned, it can usually be moved and is still trippable. If the rod can be realigned within the Completion Time of 1 hour, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction. An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour gives the operator sufficient time to adjust the rod positions in an orderly manner.

B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit. In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps. Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour. The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration. (OPERABLE)481misaligned but8 Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-6 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

ACTIONS (continued) B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (FQ(Z) and HNF) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible. Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 7). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met. Verifying that FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(Z) and HNF. Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis. C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0 within 6 hours, which 5 U 31144X,Y,11resulting from 1X,Y,1X,Y,FH(X,Y)FH(X,Y)FH(X,Y) Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-7 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

ACTIONS (continued) obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems. D.1.1 and D.1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0 within 6 hours.

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.4.1 REQUIREMENTS [ Verification that individual rod positions are within alignment limits at a Frequency of 12 hours provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected. OR 3 34459of 1of the control banks Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-8 Rev. 4.0 SEQUOYAH UNIT 2 1Revision XXX BASES

SURVEILLANCE REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2 with Keff 1.0, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. [ The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken. 6556541greater than or equal to in either direction Rod Group Alignment Limits B 3.1.4 Westinghouse STS B 3.1.4-9 Rev. 4.0 1Revision XXXSEQUOYAH UNIT 2 BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 500°F to simulate a reactor trip under actual conditions. This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. FSAR, Chapter [15].

4. FSAR, Chapter [15].

5. FSAR, Chapter [15].

6. FSAR, Chapter [15]. 7. FSAR, Chapter [15]. U Section 15.2.3Section 15.4.2U 17171711INSERT 31installation1 B 3.1.4 Insert Page B 3.1.4-9 INSERT 3 Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of 222 and 231 steps withdrawn, inclusive. 1 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4 BASES, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3. 3. ISTS B 3.1.4 Applicable Safety Analyses section contains discussion of the Required Action when the LCO is not met. ITS B 3.1.4 Applicable Safety Analyses section does not contain this discussion. This information is adequately addressed in the Bases for ACTIONS 4. Changes are made to be consistent with the Specification. 5. ISTS SR 3.1.4.1 and SR 3.1.4.2 Bases provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 6. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 7. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 8. Editorial changes made for enhanced clarity/consistency. 9. Typographical/grammatical error corrected.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 5 ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.5 REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the COLR. APPLICABILITY: MODES 1* and 2*# ACTION:

a. With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2 or when complying with ACTION b of this specification, within one hour either: 1. Restore the rod to within the insertion limit specified in the COLR, or 2. Declare the rod to be inoperable and apply ACTION 3.1.3.1.c.3.

b. With a maximum of one shutdown bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable due to malfunctions in the rod control system, POWER OPERATION may continue provided that: 1. The shutdown bank is inserted no more than 18 steps below the insertion limit as measured by the group step counter demand position indicators, 2. The affected bank is trippable, 3. Each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position, 4. The insertion limits of Specification 3.1.3.6 are met for each control bank, 5. No reactor coolant system boron concentration dilution activities or power level increases are allowed, 6. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined to be met at least once per 12 hours or upon insertion of the controlling bank more than 5 steps from the initial position, and 7. The shutdown bank is restored to within the insertion limit specified in the COLR within 72 hours.

Otherwise, be in HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLR: a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and b. At least once per 12 hours thereafter. ________________ *See Special Test Exceptions 3.10.2 and 3.10.3. #With Keff greater than or equal to 1.0. November 21, 1995 SEQUOYAH - UNIT 1 3/4 1-20 Amendment No. 108, 155, 215 A03M01twoPage 1 of 2 LCO 3.1.5 Applicability ACTION C Applicability Note ACTION B L02A03M01In accordance with the Surveillance Frequency Control Program LA01ACTION A SR 3.1.5.1 Add proposed ACTION C L01one or more shutdown banksBANK bank Each A02A05A04Each control and shutdown rod within the limits of LCO 3.1.4 Add proposed Required Actions B.1.1 and B.1.2bank bank A02 A01ITS ITS 3.1.5 REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the COLR: APPLICABILITY: Modes 1* and 2*#. ACTION:

a. With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2 or when complying with ACTION b of this specification, within one hour either:

1. Restore the rod to within the insertion limit specified in the COLR, or 2. Declare the rod to be inoperable and apply ACTION 3.1.3.1.c.3.

b. With a maximum of one shutdown bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable due to malfunctions in the rod control system, POWER OPERATION may continue provided that:

1. The shutdown bank is inserted no more than 18 steps below the insertion limit as measured by the group step counter demand position indicators, 2. The affected bank is trippable, 3. Each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position, 4. The insertion limits of Specification 3.1.3.6 are met for each control bank, 5. No reactor coolant system boron concentration dilution activities or power level increases are allowed, 6. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined to be met at least once per 12 hours or upon insertion of the controlling bank more than 5 steps from the initial position, and 7. The shutdown bank is restored to within the insertion limit specified in the COLR within 72 hours.

Otherwise, be in HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLR: a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and

b. At least once per 12 hours thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3. # With Keff greater than or equal to 1.0 November 21, 1995 SEQUOYAH - UNIT 2 3/4 1-20 Amendment No. 98, 146, 205 Page 2 of 2 A03M01Add proposed Required Actions B.1.1 and B.1.2LCO 3.1.5 Applicability ACTION C Applicability Note ACTION B L02A03M01ACTION A SR 3.1.5.1 twoIn accordance with the Surveillance Frequency Control Program LA01one or more shutdown banksL01Add proposed ACTION C A05BANK A02A02bank Each A04Each control and shutdown rod within the limits of LCO 3.1.4 bank bank A02 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 5 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.1.3.5 states "All shutdown rods shall be limited in physical insertion as specified in the COLR. Additionally, the title of CTS 3.1.3.5 is "SHUTDOWN ROD INSERTION LIMIT." ITS LCO 3.1.5 states "Each shutdown bank shall be within insertion limits specified in the COLR." Furthermore, ITS 3.1.5 title has been changed to "SHUTDOWN BANK INSERTION LIMIT." This changes the CTS by referring to each bank instead of all rods. The purpose of CTS 3.1.3.5 is to ensure that sufficient negative reactivity is available to shutdown the reactor and to maintain the SDM. This change is acceptable because the requirements have not changed. ITS 3.1.5 will continue to ensure that sufficient negative reactivity is available to shutdown the reactor and to maintain the SDM. This change is a change in presentation to match the ISTS format. Therefore, this change is designated as an administrative change because it does not result in a technical change to the CTS. A03 CTS 3.1.3.5 Applicability is modified by a footnote (footnote *) which states "See Special Test Exceptions 3.10.2 and 3.10.3." ITS 3.1.5 Applicability does not contain this footnote or a reference to the Special Test Exceptions. This changes the CTS by not including footnote *. The purpose of Footnote

is to alert the Technical Specification user that a Special Test Exception exists that may modify the Applicability of this Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative because it does not result in a technical change to the CTS.

A04 CTS 3.1.3.5 ACTION b states that POWER OPERATION may continue with a maximum of one shutdown bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable due to malfunctions in the rod control system. ITS 3.1.5 ACTION A allows POWER OPERATION to continue with one shutdown bank inserted beyond the insertion limit and immovable due to malfunctions in the rod control system. This changes the CTS by removing the qualification statement "during surveillance testing pursuant to Specification 4.1.3.1.2." The purpose of CTS 3.1.3.5 ACTION b is to allow time for diagnosis and repair of an inoperable shutdown bank if the failure is external to the control rod drive mechanism. Since the shutdown banks are required to be fully withdrawn in DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 2 of 5 MODES 1 and 2, the only time the shutdown banks are inserted, in these MODES, are during the performance of the rod freedom of movement test of CTS 4.1.3.1.2 and low power physics testing. Therefore, the statement "during surveillance testing pursuant to Specification 4.1.3.1.2" is not necessary. Furthermore, ITS LCO 3.1.5 is not applicable during the rod freedom of movement test, as stated in the ITS 3.1.5 Applicability Note. Therefore, referencing the SR (ITS SR 3.1.4.2) within the Specification would be confusing. This change is designated as administrative because it does not result in a technical change to the specifications. A05 CTS 3.1.3.5 ACTION b states, in part, that with a maximum of one shutdown bank inserted beyond the insertion limit, POWER OPERATION may continue provided that the affected bank is trippable and each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position. ITS 3.1.5 Required Action A.2 requires immediate verification that each control and shutdown rod are within the limits of LCO 3.1.4. This changes the CTS by specifically stating that the control and shutdown banks shall be within the limits of LCO 3.1.4. The purpose of this portion of CTS 3.1.3.5 ACTION b is to verify the requirements of CTS 3.1.3.1 are met. CTS 3.1.3.1 states that all full length (shutdown and control) rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position. In CTS 3.1.3.5 ACTION b, verifying that the affected bank is trippable, is verifying that the bank is OPERABLE. Additionally, verifying each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position in CTS 3.1.3.5, is the same as verifying the shutdown and control rods are positioned within +/- 12 steps (indicated position) of their group step counter demand position. The ITS 3.1.5 Required Action B.2 statement eliminates any confusion as to what actions are being taken. This change is designated as administrative because it does not result in a technical change to the specifications.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.3.5 is applicable in MODES 1 and 2 with keff 1.0. MODE 2 is modified by CTS 3.1.3.5 footnote #. ITS 3.1.5 is applicable in MODES 1 and 2. This changes the CTS by expanding the Applicability from MODE 2 with the reactor critical to all of MODE 2. The purpose of CTS 3.1.3.5 is to ensure that the shutdown banks are fully withdrawn prior to withdrawing the control banks in order to ensure that there is sufficient shutdown margin available to quickly shutdown the reactor. This change is acceptable because applying the requirement prior to removing the control banks and bringing the reactor critical ensures that the shutdown margin is available and is consistent with plant operation, in that the shut down banks are completely withdrawn before beginning to withdraw the control banks and approaching criticality. This change is designated as more restrictive because it DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 3 of 5 increases the conditions under which Technical Specification controls will be applied.

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.1.3.5.b requires verification that each shutdown rod is within the insertion limit specified in the COLR at least once per 12 hours. ITS 3.1.5.1 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program.

The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.3.5 ACTION a provides compensatory actions for a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR. The actions require within one hour either restore the rod to within the insertion limit specified in the COLR or declare the rod to be inoperable and apply ACTION 3.1.3.1.c.3. For more than one shutdown rod beyond the insertion limit, CTS 3.1.3.5 does not contain a specific requirement; therefore, entry into CTS 3.0.3 is required. ITS 3.1.5 ACTION B provides Required Actions for one or more shutdown banks not within limits. ITS 3.1.5 Required Action B.1 requires either verification the SDM is within the limits specified in the COLR (Required Action B.1.1) or the initiation of boration to restore SDM to within limits (Required Action B.1.2), both within 1 hour. ITS 3.1.5 Required Action B.2 requires restoration of the shutdown banks to DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 4 of 5 within limits within 2 hours. Additionally, ITS 3.1.5 ACTION C requires if any Required Action and associated Completion Time is not met, the unit must be in MODE 3 within 6 hours. This changes the CTS by allowing more than one shutdown rod to be outside the insertion limits specified in the COLR, provides an additional hour to restore the shutdown bank or shutdown rod to within limits, eliminates the allowance to declare the rod inoperable and to take the ACTIONS of Specification 3.1.3.1, and adds the requirement to verify SDM or to initiate boration within one hour. It also eliminates the requirement to enter CTS 3.0.3 if more than one shutdown rod is inserted beyond the insertion limits. The purpose of CTS 3.1.3.5 ACTION a is to ensure the shutdown banks are fully withdrawn in order to ensure that there is sufficient margin available to quickly shutdown the reactor. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering that only a small amount of time is provided to establish the required features, and the low probability of a DBA occurring during the repair period. Allowing an additional hour to restore one or more shutdown banks (or more than one shutdown rod) inserted below the insertion limit is appropriate as it may avoid a shutdown, a unit transient, while the rod control system is not in full working order. The ITS requires verification that the shutdown margin requirement is met or actions to restore the shutdown margin to within its limit within 1 hour, so all safety analysis assumptions are being met. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L02 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.3.5.a requires verification that each shutdown rod is within the insertion limit specified in the COLR within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor criticality. ITS 3.1.5 does not require verification that the shutdown rods are above the insertion limits within 15 minutes prior to control bank withdrawal. This changes the CTS by eliminating the requirement that the shutdown banks be verified to be above the insertion limit within 15 minutes prior to withdrawing control banks A, B, C, and D. The purpose of CTS 4.1.3.5.a is to verify the shutdown rods are withdrawn above the insertion limit prior to withdrawing the control banks. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify the equipment being used to meet the LCO can perform its required function. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence the equipment can perform its assumed safety function. Under the ITS Applicability of MODE 2 and the requirement of ITS LCO 3.0.4, the shutdown banks must be above the insertion limit prior to entering the ITS Applicability of MODE 2. However, it is not required to verify compliance within a specified time prior to initial control bank withdrawal. Specifying a time is not necessary to ensure the shutdown banks are above the insertion limit prior to initial control bank withdrawal as long as the shutdown banks are withdrawn before withdrawing the control banks. This change is DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 5 of 5 designated as less restrictive because a Surveillance which was required in CTS will not be required in the ITS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Shutdown Bank Insertion Limits 3.1.5 Westinghouse STS 3.1.5-1 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX33.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limits

LCO 3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR. APPLICABILITY: MODES 1 and 2. ------------------------------------------NOTE----------------------------------------------- This LCO is not applicable while performing SR 3.1.4.2.

  --------------------------------------------------------------------------------------------------

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown banks not within limits. A.1.1 Verify SDM is within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Restore shutdown banks to within limits. 1 hour

1 hour

2 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. 6 hours INSERT 1C C 3.1.3.5 Applicability ACTION a ACTION a ACTION b, DOC L01 for reasons other than Condition A 11B B B B 11(s)2 3.1.5 Insert Page 3.1.5-1 CTS INSERT 1 A. One shutdown bank not within limits and immovable due to malfunctions in the Rod Control System. A.1 Verify shutdown bank is inserted 18 steps below the insertion limit as measured by group step counter demand position indicators. AND A.2 Verify each control and shutdown rod is within limits of LCO 3.1.4, "Rod Group Alignment Limits." AND A.3 Verify each control bank is within insertion limits of LCO 3.1.6, "Rod Group Insertion Limits.". AND A.4 Verify no Reactor Coolant System boron dilution activities in progress. AND A.5 Verify no power level increases. AND A.6 Verify SDM is within limits specified in the COLR.

AND A.7 Restore shutdown bank to within limits. Immediately

Immediately

Once per 12 hours AND Immediately upon insertion of controlling bank more than 5 steps from the initial position

72 hours ACTION b 1 Shutdown Bank Insertion Limits 3.1.5 Westinghouse STS 3.1.5-2 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the insertion limits specified in the COLR. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 4.1.3.5 44 Shutdown Bank Insertion Limits 3.1.5 Westinghouse STS 3.1.5-1 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX33.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limits

LCO 3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR. APPLICABILITY: MODES 1 and 2. ------------------------------------------NOTE----------------------------------------------- This LCO is not applicable while performing SR 3.1.4.2.

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ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown banks not within limits. A.1.1 Verify SDM is within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Restore shutdown banks to within limits. 1 hour

1 hour

2 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. 6 hours INSERT 1C C 3.1.3.5 Applicability ACTION a ACTION a ACTION b, DOC L01 for reasons other than Condition A 11B B B B 11(s)2 3.1.5 Insert Page 3.1.5-1 CTS INSERT 1 A. One shutdown bank not within limits and immovable due to malfunctions in the Rod Control System. A.1 Verify shutdown bank is inserted 18 steps below the insertion limit as measured by group step counter demand position indicators. AND A.2 Verify each control and shutdown rod is within limits of LCO 3.1.4, "Rod Group Alignment Limits." AND A.3 Verify each control bank is within insertion limits of LCO 3.1.6, "Rod Group Insertion Limits.". AND A.4 Verify no Reactor Coolant System boron dilution activities in progress. AND A.5 Verify no power level increases. AND A.6 Verify SDM is within limits specified in the COLR.

AND A.7 Restore shutdown bank to within limits. Immediately

Immediately

Once per 12 hours AND Immediately upon insertion of controlling bank more than 5 steps from the initial position

72 hours ACTION b 1 Shutdown Bank Insertion Limits 3.1.5 Westinghouse STS 3.1.5-2 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the insertion limits specified in the COLR. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 4.1.3.5 44 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. ISTS 3.1.5 has been modified to include a new ACTION (ITS 3.1.5 ACTION A). ITS 3.1.5 requires entering Condition A when one shutdown bank is inserted beyond the insertion limit and immovable due to a malfunction in the rod control system. ITS 3.1.5 Required Action A.1 requires an immediate verification that the shutdown bank is inserted less than or equal to 18 steps below the insertion limit as measured by the group step counter demand position indicators. ITS 3.1.5 Required Action A.2 requires an immediate verification that each control and shutdown rod is within the limits of LCO 3.1.4. ITS 3.1.5 Required Action A.3 requires an immediate verification that each control bank is within the insertion limits of LCO 3.1.6. ITS 3.1.5 Required Action A.4 requires an immediate verification that there are no reactor coolant system boron dilution concentration activities. ITS 3.1.5 Required Action A.5 requires an immediate verification that there are no power level increases. ITS 3.1.5 Required Action A.6 requires verification that the SDM is within the limits specified in the COLR once per 12 hours and upon insertion of the controlling bank more than 5 steps from the initial position. ITS 3.1.5 Required Action A.7 requires the restoration of the shutdown bank to within limits in 72 hours. This addition is acceptable because it reflects the current licensing basis. Furthermore, ISTS 3.1.5 Condition A (ITS 3.1.5 Condition B) was modified to state it is applicable for reasons other than Condition A, consistent with current licensing. This change was approved in License Amendment 215 for Unit 1 and License Amendment 205 for Unit 2 (ADAMS Accession No. ML013330266). Additionally, due to the addition of ITS 3.1.5 ACTION A, the subsequent ACTIONS were renumbered. 2. Editorial changes made for enhanced clarity/consistency.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 4. ISTS SR 3.1.5.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-1 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.5 Shutdown Bank Insertion Limits

BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements. The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally automatically controlled by the Rod Control System, but they can also be manually controlled. They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operations.

Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The Each unit has 1four Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-2 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

BACKGROUND (continued) shutdown banks are controlled manually by the control room operator. During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal. APPLICABLE On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFETY the most reactive RCCA, are assumed to insert into the core. The ANALYSES shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal. The control banks may be partially inserted in the core, as allowed by LCO 3.1.6, "Control Bank Insertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod. The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that: a. There be no violations of: 1. Specified acceptable fuel design limits or 2. RCS pressure boundary integrity and b. The core remains subcritical after accident transients. As such, the shutdown bank insertion limits affect safety analysis involving core reactivity and SDM (Ref. 3). The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 11INSERT 1 e 2 B 3.1.5 Insert Page B 3.1.5-2 INSERT 1 They are moved quarterly or following maintenance to ensure trippability but are returned to the withdrawn position when the testing is completed. 1 Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-3 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown bank insertion limits are defined in the COLR. APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration," ensures adequate SDM in MODE 6. The Applicability requirements have been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the shutdown bank to move below the LCO limits, which would normally violate the LCO. ACTIONS A.1.1, A.1.2, and A.2 When one or more shutdown banks is not within insertion limits, 2 hours is allowed to restore the shutdown banks to within the insertion limits. This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1. The allowed Completion Time of 2 hours provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

B.1 If the shutdown banks cannot be restored to within their insertion limits within 2 hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. C INSERT 233INSERT 3, except for control rod OPERABILITY testing, 1B B B3for reasons other than Condition AMODE 2 keff < 1.0, 3 B 3.1.5 Insert Page B 3.1.5-3 INSERT 2 A.1, A.2, A.3, A.4, A.5, A.6, and A.7 When one shutdown bank is inserted beyond the insertion limit and is immovable due to a malfunction in the rod control system, 72 hours are provided to restore the shutdown banks to within limits. Additionally, immediate verification is required to prove that the shutdown bank is less than or equal to 18 steps below the insertion limit as measured by the group demand position indicators, the individual control rod alignment limits of LCOs 3.1.4 and 3.1.6 are met, there are no reactor coolant system boron dilution activities, and there are no power level increases are taking place. Furthermore, a verification of SDM is required within 12 hours or when the controlling banks are inserted more than 5 steps from the initial position. The requirement to be in compliance with LCOs 3.1.4 and 3.1.6 ensures that the rods are trippable, and power distribution is acceptable during the time allowed to restore the inserted rod. The 12 hour requirement to verify the SDM is within limits ensures the SDM requirements of LCO 3.1.1 are met during the repair period. Furthermore, the requirement to verify the SDM is within limits when a controlling bank is inserted five steps or more also ensures that SDM requirements of LCO 3.1.1 are met during the repair period. If any of these Conditions are not met, Condition C must be applied. The Completion Time of 72 hours is based on operating experience and provides an acceptable time for evaluating and repairing problems with the rod control system. INSERT 3 the Required Action(s) of Condition A or B are not met within the associated Completion Times 22 Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-4 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup. [ Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours, after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28. 2. 10 CFR 50.46. 3. FSAR, Chapter [15]. U 45416 Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-1 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.5 Shutdown Bank Insertion Limits

BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements. The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally automatically controlled by the Rod Control System, but they can also be manually controlled. They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operations.

Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The Each unit has 1four Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-2Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES BACKGROUND (continued) shutdown banks are controlled manually by the control room operator. During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal. APPLICABLE On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFETY the most reactive RCCA, are assumed to insert into the core. The ANALYSES shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal. The control banks may be partially inserted in the core, as allowed by LCO 3.1.6, "Control Bank Insertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod. The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that: a.There be no violations of:1.Specified acceptable fuel design limits or2.RCS pressure boundary integrity andb.The core remains subcritical after accident transients.As such, the shutdown bank insertion limits affect safety analysis involving core reactivity and SDM (Ref. 3). The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 11INSERT 1 e 2 B 3.1.5 Insert Page B 3.1.5-2 INSERT 1 They are moved quarterly or following maintenance to ensure trippability but are returned to the withdrawn position when the testing is completed. 1 Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-3Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown bank insertion limits are defined in the COLR. APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration," ensures adequate SDM in MODE 6. The Applicability requirements have been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the shutdown bank to move below the LCO limits, which would normally violate the LCO. ACTIONS A.1.1, A.1.2, and A.2 When one or more shutdown banks is not within insertion limits, 2 hours is allowed to restore the shutdown banks to within the insertion limits. This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1. The allowed Completion Time of 2 hours provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time. B.1 If the shutdown banks cannot be restored to within their insertion limits within 2 hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. C INSERT 233INSERT 3, except for control rod OPERABILITY testing, 1B B B3for reasons other than Condition AMODE 2 keff < 1.0, 3 B 3.1.5 Insert Page B 3.1.5-3 INSERT 2 A.1, A.2, A.3, A.4, A.5, A.6, and A.7 When one shutdown bank is inserted beyond the insertion limit and is immovable due to a malfunction in the rod control system, 72 hours are provided to restore the shutdown banks to within limits. Additionally, immediate verification is required to prove that the shutdown bank is less than or equal to 18 steps below the insertion limit as measured by the group demand position indicators, the individual control rod alignment limits of LCOs 3.1.4 and 3.1.6 are met, there are no reactor coolant system boron dilution activities, and there are no power level increases are taking place. Furthermore, a verification of SDM is required within 12 hours or when the controlling banks are inserted more than 5 steps from the initial position. The requirement to be in compliance with LCOs 3.1.4 and 3.1.6 ensures that the rods are trippable, and power distribution is acceptable during the time allowed to restore the inserted rod. The 12 hour requirement to verify the SDM is within limits ensures the SDM requirements of LCO 3.1.1 are met during the repair period. Furthermore, the requirement to verify the SDM is within limits when a controlling bank is inserted five steps or more also ensures that SDM requirements of LCO 3.1.1 are met during the repair period. If any of these Conditions are not met, Condition C must be applied. The Completion Time of 72 hours is based on operating experience and provides an acceptable time for evaluating and repairing problems with the rod control system. INSERT 3 the Required Action(s) of Condition A or B are not met within the associated Completion Times 22 Shutdown Bank Insertion Limits B 3.1.5 Westinghouse STS B 3.1.5-4 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup. [ Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours, after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28. 2. 10 CFR 50.46. 3. FSAR, Chapter [15]. U 45416 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5 BASES, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. Editorial changes made for enhanced clarity/consistency. 3. Changes are made to be consistent with changes made to the Specification. Additionally, the subsequent ACTIONS have been renumbered. 4. ISTS SR 3.1.5.1 and SR 3.1.5.2 Bases provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 6. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 6 ITS 3.1.6, CONTROL BANK INSERTION LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.6 REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the COLR. APPLICABILITY: MODES 1* and 2*#. ACTION:

a. With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2 or when complying with ACTION b of this specification, either: 1. Restore the control banks to within the limits within two hours, or 2. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the COLR, or 3. Be in HOT STANDBY within 6 hours.

b. With a maximum of one control bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable due to malfunctions in the rod control system, POWER OPERATION## may continue provided that: 1. The control bank is inserted no more than 18 steps below the insertion limit as measured by the group step counter demand position indicators, 2. The affected bank is trippable, 3. Each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position, 4. The insertion limits of Specification 3.1.3.5 are met for each shutdown bank, 5. No reactor coolant system boron concentration dilution activities or power level increases are allowed, 6. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined to be met at least once per 12 hours or upon insertion of the controlling bank more than 5 steps from the initial position, and 7. The control bank is restored to within the insertion limit specified in the COLR within 72 hours. Otherwise, be in HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours. _________________ *See Special Test Exceptions 3.10.2 and 3.10.3. #With Keff greater than or equal to 1.0.

1. Provision for continued POWER OPERATION does not apply to the controlling bank(s) (normally Control Bank D) inserted beyond the insertion limit. November 21, 1995 SEQUOYAH - UNIT 1 3/4 1-21 Amendment No. 41, 114, 155, 215 LCO 3.1.6 Page 1 of 8 M02Add proposed Required Action B.1.1 and B.1.2 , sequence, and overlap limits M01A02A03M01Add proposed ACTION C MODE 2 with keff < 1.0 A04M01Add proposed SR 3.1.6.3Applicability ACTION D SR 3.1.6.2 Applicability Note ACTION B ACTION A MODE 2 with keff < 1.0 A04L01A06Each control and shutdown rod within the limits of LCO 3.1.4. ACTION A Note A02Applicability ACTION D A01BANK A05In accordance with the Surveillance Frequency Control Program LA01LA02 A01ITS ITS 3.1.6

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October 23, 1991 SEQUOYAH - UNIT 1 3/4 1-22 Amendment No. 108, 155 Page 2 of 8 A01ITS ITS 3.1.6

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October 23, 1991 SEQUOYAH - UNIT 1 3/4 1-23 Amendment No. 41, 108, 155 Page 3 of 8 A01ITS ITS 3.1.6 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg Greater Than 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s). b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

____________________ *See Special Test Exception 3.10.1 November 26, 1993 SEQUOYAH - UNIT 1 3/4 1-1 Amendment No. 172 Page 4 of 8 See ITS 3.1.1 See ITS 3.1.4 See ITS 3.1.1 See ITS Chapter 1.0SR 3.1.6.1 SR 3.1.6.2 In accordance with the Surveillance Frequency Control Program LA01See ITS 3.1.1 See ITS 3.1.1 Applicability A01ITS ITS 3.1.6 REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the COLR APPLICABILITY: Modes 1* and 2*#. ACTION: a. With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2 or when complying with ACTION b of this specification, either: 1. Restore the control banks to within the limits within two hours, or 2. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the COLR, or 3. Be in HOT STANDBY within 6 hours.

b. With a maximum of one control bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable due to malfunctions in the rod control system, POWER OPERATION## may continue provided that: 1. The control bank is inserted no more than 18 steps below the insertion limit as measured by the group step counter demand position indicators, 2. The affected bank is trippable, 3. Each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position, 4. The insertion limits of Specification 3.1.3.5 are met for each shutdown bank, 5. No reactor coolant system boron concentration dilution activities or power level increases are allowed, 6. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined to be met at least once per 12 hours or upon insertion of the controlling bank more than 5 steps from the initial position, and 7. The control bank is restored to within the insertion limit specified in the COLR within 72 hours. Otherwise, be in HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours.

See Special Test Exceptions 3.10.2 and 3.10.3. # With Keff greater than or equal to 1.0. ## Provision for continued POWER OPERATION does not apply to the controlling bank(s) (normally Control Bank D) inserted beyond the insertion limit.

November 21, 1995 SEQUOYAH - UNIT 2 3/4 1-21 Amendment No. 33, 104, 146, 205 LCO 3.1.6 Page 5 of 8 M02Add proposed Required Action B.1.1 and B.1.2 , sequence, and overlap limits M01A02A03M01Add proposed ACTION C MODE 2 with keff < 1.0 A04M01Add proposed SR 3.1.6.3Applicability ACTION D SR 3.1.6.2 Applicability Note ACTION B ACTION A MODE 2 with keff < 1.0 A04L01A06Each control and shutdown rod within the limits of LCO 3.1.4 ACTION A Note LA01In accordance with the Surveillance Frequency Control Program A02Applicability A01BANK A05LA02 A01ITS ITS 3.1.6

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March 30, 1992 SEQUOYAH - UNIT 2 3/4 1-22 Amendment Nos. 98, 146 Page 6 of 8 A01ITS ITS 3.1.6

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March 30, 1992 SEQUOYAH - UNIT 2 3/4 1-23 Amendment No. 33, 98, 146 Page 7 of 8 A01ITS ITS 3.1.6 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k for 4 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2, with Keff less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1

November 26, 1993 SEQUOYAH - UNIT 2 3/4 1-1 Amendment No. 163 Page 8 of 8 See ITS 3.1.1 See ITS 3.1.4 See ITS 3.1.1 See ITS Chapter 1.0SR 3.1.6.1 SR 3.1.6.2 In accordance with the Surveillance Frequency Control Program LA01See ITS 3.1.1 See ITS 3.1.1 See ITS 3.1.1 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 5 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.1.3.6 Applicability is modified by a footnote (footnote *) that states "See Special Test Exceptions 3.10.2 and 3.10.3." ITS 3.1.6 Applicability does not contain the footnote or a reference to the Special Test Exceptions. This changes the CTS by not including footnote *. The purpose of Footnote

is to alert the Technical Specification user that a Special Test Exception exists that may modify the Applicability of this Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 3.1.3.6 ACTION a states that with the control banks beyond the insertion limits, to restore the control bank to within limits within 2 hours or reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the COLR. ITS 3.1.6 Required Action B.2 requires restoring the control banks to within limits within 2 hours. This changes the CTS by eliminating the explicit statement that compliance with the LCO can be restored in order to exit the ACTION. This change is acceptable because the requirements have not changed. When THERMAL POWER is reduced, the insertion limits, which are a function of power, are lowered. When the insertion limits are lowered, the control banks, which were previously inserted below the insertion limits, will then come within the new limit. This is the same as the CTS ACTION a option to restore the control banks to within the limit. This change is considered administrative because the technical requirements have not changed.

A04 CTS 3.1.3.6 ACTION a.3 and ACTION b require the unit to be in HOT STANDBY (MODE 3) within 6 hours if ACTION a or b are not met. The CTS Applicability is MODES 1 and 2 with keff 1.0. ITS 3.1.6 ACTION D requires the unit to be in MODE 2 with keff < 1.0. This changes the CTS by requiring the unit to be in MODE 2 with keff < 1.0 instead of HOT STANDBY (MODE 3). This change is acceptable because the requirements have not changed. In the CTS, ACTIONS are only required to be followed while in the Mode of Applicability. The CTS control bank insertion limits are applicable in MODES 1 and 2 with keff 1.0. Therefore, under the CTS, the unit does not have to enter DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 2 of 5 MODE 3 because the Applicability of the LCO has been exited when in MODE 2 with keff < 1.0. As a result, there is no difference between the CTS and the ITS requirements. This change is designated as administrative because it does not result in a technical change to the CTS. A05 CTS 3.1.3.6 ACTION b states that POWER OPERATION may continue with a maximum of one control bank inserted beyond the insertion limit specified in the COLR during surveillance testing pursuant to Specification 4.1.3.1.2 and immovable resulting from malfunctions in the rod control system. ITS 3.1.6 ACTION A allows, in part, POWER OPERATION to continue with one control bank inserted beyond the insertion limit and immovable. This changes the CTS by removing the qualification statement "during surveillance testing pursuant to Specification 4.1.3.1.2." The purpose of CTS 3.1.3.6 ACTION b is to allow time for diagnosis and repair to an inoperable control bank if the failure is external to the control rod drive mechanism. Since the shutdown banks are required to be fully withdrawn in MODES 1 and 2, the only time the control banks are inserted, in these MODES, are during the performance of the rod freedom test of CTS 4.1.3.1.2. Therefore, the statement "during surveillance testing pursuant to Specification 4.1.3.1.2" is not necessary. Furthermore, ITS LCO 3.1.6 is not applicable during the rod freedom test, as stated in the ITS 3.1.6 Applicability Note. Therefore, referencing the SR (ITS SR 3.1.4.2) within the Specification would be confusing. This change is designated as administrative because it does not result in a technical change to the specifications. A06 CTS 3.1.3.6 ACTION b states, in part, that with a maximum of one control bank inserted beyond the insertion limit, POWER OPERATION may continue provided that the affected bank is trippable and each shutdown and control rod is aligned to within +/- 12 steps of its respective group step counter demand position. ITS 3.1.6 Required Action A.2 requires, in part, verification that each control and shutdown rod is within the limits of LCO 3.1.4. This changes the CTS by specifically stating that the control and shutdown rods shall be verified to be within the limits of LCO 3.1.4. The purpose of this portion of CTS 3.1.3.6 ACTION b is to verify the requirements of CTS 3.1.3.1 are met. CTS 3.1.3.1 states that all full length (shutdown and control) rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position. In CTS 3.1.3.6 ACTION b, verifying that the affected bank is trippable, is verifying that the bank is OPERABLE. Additionally, when the control rod is aligned to within +/- 12 steps of its respective group step counter demand position in CTS 3.1.3.6, this is the same as verifying the shutdown and control rods are positioned within +/- 12 steps (indicated position) of their group step counter demand position. The ITS 3.1.6 Required Action A.2 statement eliminates any confusion as to what actions are being taken. This change is designated as administrative because it does not result in a technical change to the specifications. DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 3 of 5 MORE RESTRICTIVE CHANGES M01 CTS 3.1.3.6 requires the control banks to be limited in physical insertion as specified in the COLR. ITS LCO 3.1.6 requires the control banks to be within insertion, sequence and overlap limits specified in the COLR. ITS 3.1.6 ACTION C provides requirements when not meeting the sequence and overlap requirements. ITS SR 3.1.6.3 requires verification of the sequence and overlap limits every 12 hours. This changes the CTS by adding the requirements on the sequence and overlap limits in addition to the Technical Specifications. This change is acceptable because the control bank sequence and overlap limits are important assumptions in the core power distribution analyses. The addition of these requirements, ACTIONS, and Surveillance Requirements provides assurance that the core power distribution is maintained within the design predictions. This change is designated as more restrictive because new requirements are added to the CTS. M02 CTS 3.1.3.6 ACTION a requires, in part, control banks inserted beyond the insertion limits to be restored within 2 hours. ITS 3.1.6 ACTION B contains the same requirements and adds the requirement to either verify the SDM is within limits or initiate boration to restore SDM to within limits within one hour. This changes the CTS by adding the requirement to verify SDM or to initiate boration to restore the SDM within one hour when control banks are below the insertion limits. This change is acceptable because it verifies that the initial conditions of the accident analyses are maintained. In MODE 1 and MODE 2 with keff 1.0, SDM is ensured by adhering to the control and shutdown bank insertion limits. If the control banks are not within their insertion limits, then SDM must be verified to be within limits or actions must be initiated to restore SDM to within limits. This change is designated as more restrictive because requirements are added to the CTS. RELOCATED SPECIFICATIONS None

DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 4 of 5 REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.1.3.6 requires, in part, the position of each control bank shall be determined to be within the insertion limits at least once per 12 hours. CTS 4.1.1.1.1.b requires, in part, verifying the control bank withdrawal is within limits of Specification 3.1.3.6 at least once per 12 hours. ITS SR 3.1.6.2 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.3.6 requires the control banks to be limited in physical insertion as specified in the COLR. CTS 3.1.3.6 ACTION b allows POWER OPERATION to continue with a maximum of one control bank inserted beyond the limit specified in the COLR during the rod freedom of movement surveillance provided the control bank is immovable due to a malfunction of the rod control system and the specified actions are met within the specified times specified. Additionally, footnote ## states the provision for continued POWER OPERATION does not apply to the controlling bank(s) (normally Control Bank D) inserted beyond the insertion limit. ITS LCO 3.1.6 and ACTION A retain the same requirements, but do not specify that Control Bank D is normally the controlling bank. This changes the CTS by relocating the details that Control Bank D is normally the controlling bank to the Bases. The removal of these details, that are related to system design, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement for the control banks to be within the insertion limits specified in the COLR, as well as the Actions to take when a control bank is not within the limits specified in the COLR. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 5 of 5 the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1.3.6 requires verification that each control rod is within the insertion limit at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then it requires verification of the individual rod positions at least once per 4 hours. ITS 3.1.6.2 requires verification that each control bank insertion is within the insertion limits specified in the COLR in accordance with the Surveillance Frequency Control Program. This changes the CTS by eliminating the requirement to verify the control bank insertion to be within limits every 4 hours when the Rod Insertion Limit Monitor is inoperable. The purpose of CTS 4.1.3.6 is to periodically verify that the rods are within the alignment limit specified in the LCO. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of rod position verification when the Rod Insertion Limit Monitor is inoperable is unnecessary because inoperability of the alarm does not increase the possibility that the control banks are inserted below the limits. The Rod Insertion Limit Monitor alarm is for indication only; its use is not credited in any of the safety analyses. This change is designated as less restrictive because a Surveillance which was required in CTS will not be required in the ITS.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Control Bank Insertion Limits 3.1.6 Westinghouse STS 3.1.6-1 Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Control Bank Insertion Limits

LCO 3.1.6 Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR. APPLICABILITY: MODE 1, MODE 2 with keff 1.0. -------------------------------------------NOTE---------------------------------------------- This LCO is not applicable while performing SR 3.1.4.2.

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ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control bank insertion limits not met. A.1.1 Verify SDM is within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Restore control bank(s) to within limits. 1 hour

1 hour

2 hours B. Control bank sequence or overlap limits not met. B.1.1 Verify SDM is within the limits specified in the COLR. OR B.1.2 Initiate boration to restore SDM to within limit. AND 1 hour

1 hour for reasons other than Condition A INSERT 13.1.3.6 Applicability, Footnote # ACTION a ACTION a DOC M01 11B B B B C C C 1111 3.1.6 Insert Page 3.1.6-1 CTS INSERT 1 A. ------------NOTE------------ Only applicable to control bank(s) that are not a controlling bank. --------------------------------- One control bank not within limits and immovable due to malfunctions in the Rod Control System. A.1 Verify control bank is inserted 18 steps below the insertion limit as measured by group step demand position indicators. AND A.2 Verify each control and shutdown rod is within limits of LCO 3.1.4, "Rod Group Alignment Limits." AND A.3 Verify each shutdown bank is within insertion limits of LCO 3.1.5, "Shutdown Bank Insertion Limits." AND A.4 Verify no Reactor Coolant System boron dilution activities. AND A.5 Verify no power level increases. AND A.6 Verify SDM is within limits specified in the COLR.

AND A.7 Restore control bank to within limits. Immediately

Immediately

Once per 12 hours AND Immediately upon insertion of controlling bank more than 5 steps from the initial position

72 hours ACTION b 1 Control Bank Insertion Limits 3.1.6 Westinghouse STS 3.1.6-2 Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2 Restore control bank sequence and overlap to within limits. 2 hours C. Required Action and associated Completion Time not met. C.1 Be in MODE 2 with keff < 1.0. 6 hours

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within the limits specified in the COLR. Within 4 hours prior to achieving criticality

SR 3.1.6.2 Verify each control bank insertion is within the insertion limits specified in the COLR. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] D D DOC M01 ACTION a.3, ACTION b 4.1.1.1.1.c 4.1.3.6, 4.1.1.1.1.b DOC M01 13333C 1 Control Bank Insertion Limits 3.1.6 Westinghouse STS 3.1.6-1 Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Control Bank Insertion Limits

LCO 3.1.6 Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR. APPLICABILITY: MODE 1, MODE 2 with keff 1.0. -------------------------------------------NOTE---------------------------------------------- This LCO is not applicable while performing SR 3.1.4.2.

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ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control bank insertion limits not met. A.1.1 Verify SDM is within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Restore control bank(s) to within limits. 1 hour

1 hour

2 hours B. Control bank sequence or overlap limits not met. B.1.1 Verify SDM is within the limits specified in the COLR. OR B.1.2 Initiate boration to restore SDM to within limit. AND 1 hour

1 hour for reasons other than Condition A INSERT 13.1.3.6 Applicability, Footnote # ACTION a ACTION a DOC M01 11B B B B C C C 1111 3.1.6 Insert Page 3.1.6-1 CTS INSERT 1 A. ------------NOTE------------ Only applicable to control bank(s) that are not a controlling bank. --------------------------------- One control bank not within limits and immovable due to malfunctions in the Rod Control System. A.1 Verify control bank is inserted 18 steps below the insertion limit as measured by group step demand position indicators. AND A.2 Verify each control and shutdown rod is within limits of LCO 3.1.4, "Rod Group Alignment Limits." AND A.3 Verify each shutdown bank is within insertion limits of LCO 3.1.5, "Shutdown Bank Insertion Limits." AND A.4 Verify no Reactor Coolant System boron dilution activities. AND A.5 Verify no power level increases. AND A.6 Verify SDM is within limits specified in the COLR.

AND A.7 Restore control bank to within limits. Immediately

Immediately

Once per 12 hours AND Immediately upon insertion of controlling bank more than 5 steps from the initial position

72 hours ACTION b 1 Control Bank Insertion Limits 3.1.6 Westinghouse STS 3.1.6-2 Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2 Restore control bank sequence and overlap to within limits. 2 hours C. Required Action and associated Completion Time not met. C.1 Be in MODE 2 with keff < 1.0. 6 hours

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within the limits specified in the COLR. Within 4 hours prior to achieving criticality

SR 3.1.6.2 Verify each control bank insertion is within the insertion limits specified in the COLR. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] D D DOC M01 ACTION a.3, ACTION b 4.1.1.1.1.c 4.1.3.6, 4.1.1.1.1.b DOC M01 13333C 1 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. ISTS 3.1.6 has been modified to include a new ACTION (ITS 3.1.6 ACTION A). ITS 3.1.6 requires entering Condition A when one control bank is inserted beyond the insertion limit and immovable. ITS 3.1.6 Required Action A.1 requires an immediate verification that the control bank is inserted less than or equal to 18 steps below the insertion limit as measured by the group step counter demand position indicators. ITS 3.1.5 Required Action A.2 requires an immediate verification that each control and shutdown rod is within the limits of LCO 3.1.4. ITS 3.1.5 Required Action A.3 requires an immediate verification that each shutdown bank is within the insertion limits of LCO 3.1.5. ITS 3.1.5 Required Action A.4 requires an immediate verification that there are no reactor coolant system boron concentration activities. ITS 3.1.5 Required Action A.5 requires an immediate verification that there are no power level increases. ITS 3.1.6 Required Action A.6 requires verification that the SDM is within the limits specified in the COLR once per 12 hours and upon insertion of the controlling bank more than 5 steps from the initial position. ITS 3.1.6 Required Action A.7 requires the restoration of the shutdown banks to within limits in 72 hours. This addition is acceptable because it reflects the current licensing basis. Furthermore, ISTS 3.1.6 Condition A (ITS 3.1.6 Condition B) was modified to state it is applicable for reasons other than Condition A, consistent with current licensing. This change was approved in License Amendment 215 for Unit 1 and License Amendment 205 for Unit 2 (ADAMS Accession No. ML013330266). Additionally, due to the addition of ITS 3.1.6 ACTION A, the subsequent ACTIONS (ISTS 3.1.5 ACTIONS A, B, and C) were renumbered.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS SR 3.1.6.2 and SR 3.1.6.3 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.6 Control Bank Insertion Limits

BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements. The control bank insertion limits are specified in the COLR. An example is provided for information only in Figure B 3.1.6-1. The control banks are required to be at or above the insertion limit lines. Figure B 3.1.6-1 also indicates how the control banks are moved in an overlap pattern. Overlap is the distance travelled together by two control banks. The predetermined position of control bank C, at which control bank D will begin to move with bank C on a withdrawal, will be at 118 steps for a fully withdrawn position of 231 steps. The fully withdrawn position is defined in the COLR. Each unithas122is shown on the COLR Figure 19the four Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

BACKGROUND (continued) The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting). The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained. Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function. APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY are required to prevent power distributions that could result in fuel ANALYSES cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function. The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that: a. There be no violations of:

1. Specified acceptable fuel design limits or

2. Reactor Coolant System pressure boundary integrity and b. The core remains subcritical after accident transients.

Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES APPLICABLE SAFETY ANALYSES (continued) As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3). The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivity is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 4). Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths. The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 5). The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii), in that they are initial conditions assumed in the safety analysis. LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion. APPLICABILITY The control bank sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2 with keff 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES. The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the control bank to move below the LCO limits, which would violate the LCO. 33113MODE 2 with keff < 1.0, 9thehas9e 9e Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

ACTIONS A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position or

b. Moving rods to be consistent with power. Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour, since the SDM in MODES 1 and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") has been upset. If control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits. Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

The allowed Completion Time of 2 hours for restoring the banks to within the insertion, sequence, and overlaps limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

C.1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed within the associated Completion Times, the plant must be brought to MODE 2 with keff < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. 5D of Condition A, B, or C are not met INSERT 1 45B B BCCC9at least() B 3.1.6 Insert Page 3.1.6-4 INSERT 1 A.1, A.2, A.3, A.4, A.5, A.6, and A.7 When one control bank is inserted beyond the insertion limit and is immovable due to malfunctions in the rod control system, 72 hours are provided to restore the control banks to within limits. Additionally, immediate verification is required to prove that the control bank is less than or equal to 18 steps below the insertion limit as measured by the group demand position indicators, the individual rod alignment limits of LCOs 3.1.4 and 3.1.5 are met, there are no reactor coolant system boron concentration dilution activities, and there are no power level increases taking place. Furthermore, a verification of SDM is required within 12 hours and when the controlling bank is inserted more than 5 steps from the initial position. The requirement to be in compliance with LCOs 3.1.4 and 3.1.5 ensures that the rods are trippable, and power distribution is acceptable during the time allowed to restore the inserted bank. The 12 hour requirement to verify the SDM is within limits ensures the SDM requirements of LCO 3.1.1 are met during the repair period. Furthermore, the requirement to verify the SDM is within limits when a controlling bank is inserted five steps or more also ensures that SDM requirements of LCO 3.1.1 are met during the repair period. If any of these Conditions are not met, Condition D must be applied. The Condition is modified by a Note that specifies it only applies to control banks inserted beyond the insertion limit that are not controlling banks. A controlling bank is defined as a control bank that is less than fully withdrawn as defined in the COLR, with the exception of fully withdrawn banks that have been inserted for the performance of SR 3.1.4.2 (rod freedom of movement Surveillance). The Completion Time of 72 hours is based on operating experience and provides an acceptable time for evaluating and repairing problems with the rod control system. 4 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 1 BASES

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 [ Verification of the control bank insertion limits at a Frequency of 12 hours is sufficient to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. [ A Frequency of 12 hours is consistent with the insertion limit check above in SR 3.1.6.2. 76666 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-6 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter [15].

4. FSAR, Chapter [15]. 5. FSAR, Chapter [15]. U 667181 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-7 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX Figure B 3.1.6 (page 1 of 1) Control Bank Insertion vs. Percent RTP 2 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-1 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.6 Control Bank Insertion Limits

BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements. The control bank insertion limits are specified in the COLR. An example is provided for information only in Figure B 3.1.6-1. The control banks are required to be at or above the insertion limit lines. Figure B 3.1.6-1 also indicates how the control banks are moved in an overlap pattern. Overlap is the distance travelled together by two control banks. The predetermined position of control bank C, at which control bank D will begin to move with bank C on a withdrawal, will be at 118 steps for a fully withdrawn position of 231 steps. The fully withdrawn position is defined in the COLR. Each unithas122is shown on the COLR Figure 19the four Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-2 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

BACKGROUND (continued) The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting). The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained. Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function. APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY are required to prevent power distributions that could result in fuel ANALYSES cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function. The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that: a. There be no violations of:

1. Specified acceptable fuel design limits or

2. Reactor Coolant System pressure boundary integrity and b. The core remains subcritical after accident transients.

Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-3 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES APPLICABLE SAFETY ANALYSES (continued) As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3). The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivity is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 4). Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths. The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 5). The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii), in that they are initial conditions assumed in the safety analysis. LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion. APPLICABILITY The control bank sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2 with keff 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES. The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the control bank to move below the LCO limits, which would violate the LCO. 33113MODE 2 with keff < 1.0, 9thehas9e 9e Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-4 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

ACTIONS A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position or

b. Moving rods to be consistent with power. Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour, since the SDM in MODES 1 and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") has been upset. If control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits. Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

The allowed Completion Time of 2 hours for restoring the banks to within the insertion, sequence, and overlaps limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

C.1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed within the associated Completion Times, the plant must be brought to MODE 2 with keff < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. 5D of Condition A, B, or C are not met INSERT 1 45B B BCCC9at least() B 3.1.6 Insert Page 3.1.6-4 INSERT 1 A.1, A.2, A.3, A.4, A.5, A.6, and A.7 When one control bank is inserted beyond the insertion limit and is immovable due to malfunctions in the rod control system, 72 hours are provided to restore the control banks to within limits. Additionally, immediate verification is required to prove that the control bank is less than or equal to 18 steps below the insertion limit as measured by the group demand position indicators, the individual rod alignment limits of LCOs 3.1.4 and 3.1.5 are met, there are no reactor coolant system boron concentration dilution activities, and there are no power level increases taking place. Furthermore, a verification of SDM is required within 12 hours and when the controlling bank is inserted more than 5 steps from the initial position. The requirement to be in compliance with LCOs 3.1.4 and 3.1.5 ensures that the rods are trippable, and power distribution is acceptable during the time allowed to restore the inserted bank. The 12 hour requirement to verify the SDM is within limits ensures the SDM requirements of LCO 3.1.1 are met during the repair period. Furthermore, the requirement to verify the SDM is within limits when a controlling bank is inserted five steps or more also ensures that SDM requirements of LCO 3.1.1 are met during the repair period. If any of these Conditions are not met, Condition D must be applied. The Condition is modified by a Note that specifies it only applies to control banks inserted beyond the insertion limit that are not controlling banks. A controlling bank is defined as a control bank that is less than fully withdrawn as defined in the COLR, with the exception of fully withdrawn banks that have been inserted for the performance of SR 3.1.4.2 (rod freedom of movement Surveillance). The Completion Time of 72 hours is based on operating experience and provides an acceptable time for evaluating and repairing problems with the rod control system. 4 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-5 Rev. 4.0 1Revision XXX SEQUOYAH UNIT 2 BASES

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 [ Verification of the control bank insertion limits at a Frequency of 12 hours is sufficient to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. [ A Frequency of 12 hours is consistent with the insertion limit check above in SR 3.1.6.2. 76666 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-6 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter [15].

4. FSAR, Chapter [15]. 5. FSAR, Chapter [15]. U 667181 Control Bank Insertion Limits B 3.1.6 Westinghouse STS B 3.1.6-7 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX Figure B 3.1.6 (page 1 of 1) Control Bank Insertion vs. Percent RTP 2 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6 BASES, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. ISTS 3.1.6 contains Figure B 3.1.6-1 and states that it is an example provided for information only. ITS 3.1.6 does not include Figure B 3.1.6-1. The control bank insertion limits for Sequoyah Nuclear Plant (SQN) are located in the COLR. Therefore, ISTS Figure B 3.1.6-1 and the references to the ISTS Figure B 3.1.6-1 have been deleted. 3. Changes are made to be consistent with the Specification. 4. Typographical/grammatical error corrected. 5. Changes are made to be consistent with changes made to the Specification. 6. ISTS SR 3.1.6.2 and SR 3.1.6.3 Bases provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 7. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 8. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 9. Editorial changes made for enhanced clarity/consistency. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 7 ITS 3.1.7, ROD POSITION INDICATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown and control rod position indication system and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within +/- 12 steps. APPLICABILITY: MODES 1 and 2. ACTION: a.With a maximum of one rod position indicator per bank inoperable either:1.Determine the position of the non-indicating rod(s) indirectly by the movable incoredetectors at least once per 12 hours and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination ofthe rod's position, or2.* a) Determine the position of the non-indicating rod indirectly by the movable incore detectors within 8 hours and once every 31 days thereafter and within 8 hours if rod control system parameters indicate unintended movement, and b)Review the parameters of the rod control system for indications of unintendedrod movement for the rod with an inoperable position indicator within 16 hoursand once per 8 hours thereafter, andc)Determine the position of the non-indicating rod indirectly by the movableincore detectors within 8 hours if the rod with an inoperable position indicatoris moved greater than 12 steps and prior to increasing THERMAL POWERabove 50% RATED THERMAL POWER and within 8 hours of reaching 100%RATED THERMAL POWER, or3.Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within8 hours.b.With more than one rod position indicator per bank inoperable either:1.Determine the position of the non-indicating rod(s) indirectly by the movable incoredetectors at least once per 12 hours, and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination ofthe rod's position, and_____________________________ *Rod position monitoring by Actions 2.a), 2.b), and 2.c) may only be applied to one inoperable rodposition indicator and shall only be allowed: (1) until the end of the current cycle, or (2) until an entry into MODE 5 of sufficient duration, whichever occurs first, when the repair of the inoperable rod position indication can safely be performed. Actions 2.a), 2.b), and 2.c) shall not be allowed after the plant has been in MODE 5 or other plant condition, for a sufficient period of time, in which the repair of the inoperable rod position indication could have safely been performed. December 11, 2006 SEQUOYAH - UNIT 1 3/4 1-17 Amendment No. 118, 213, 244, 315 LCO 3.1.7 Applicability ACTION A ACTION B Add proposed ACTIONS Note 1 L01Add proposed ACTION DM01LA01Page 1 of 6 Required Action A.2 Note Add proposed ACTIONS Note 2A02 A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - OPERATING 2.Place the control rods under manual control, and monitor and record Reactor CoolantSystem average temperature (Tavg) at least once per hour, and3.Restore the rod position indicators to OPERABLE status within 24 hours such that amaximum of one rod position indicator per bank is inoperable, or4.Be in HOT STANDBY within the next 6 hours.c.With a maximum of one demand position indicator per bank inoperable either:1.Verify that all rod position indicators for the affected bank are OPERABLE and thatthe most withdrawn rod and the least withdrawn rod of the bank are within amaximum of 12 steps of each other at least once per 12 hours, or2.Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within8 hours.SURVEILLANCE REQUIREMENTS 4.1.3.2 Each rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 12 steps at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours. December 11, 2006 SEQUOYAH - UNIT 1 3/4 1-17a Amendment No. 118, 213, 244 ACTION B ACTION C ACTION D Add proposed ACTION DM01M02Add proposed SR 3.1.7.1 Page 2 of 6 A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 This specification is deleted. December 18, 2000 SEQUOYAH - UNIT 1 3/4 1-18 Amendment No. 26, 264 Page 3 of 6 A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown and control rod position indication system and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within +/- 12 steps. APPLICABILITY: Modes 1 and 2. ACTION: a.With a maximum of one rod position indicator per bank inoperable either:1.Determine the position of the non-indicating rod(s) indirectly by the movable incoredetectors at least once per 12 hours and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of therod's position, or2.* a) Determine the position of the non-indicating rod indirectly by the movable incore detectors within 8 hours and once every 31 days thereafter and within 8 hours if rod control system parameters indicate unintended movement, and b)Review the parameters of the rod control system for indications of unintendedrod movement for the rod with an inoperable position indicator within 16 hoursand once per 8 hours thereafter, andc)Determine the position of the non-indicating rod indirectly by the movable incoredetectors within 8 hours if the rod with an inoperable position indicator is moved greater than 12 steps and prior to increasing THERMAL POWER above 50%RATED THERMAL POWER and within 8 hours of reaching 100% RATEDTHERMAL POWER, or3.Reduce THERMAL POWER TO less than 50% of RATED THERMAL POWER within 8hours.b.With more than one rod position indicator per bank inoperable either:1.Determine the position of the non-indicating rod(s) indirectly by the movable incoredetectors at least once per 12 hours, and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of therod's position, and*Rod position monitoring by Actions 2.a), 2.b), and 2.c) may only be applied to one inoperable rodposition indicator and shall only be allowed: (1) until the end of the current cycle, or (2) until an entry into MODE 5 of sufficient duration, whichever occurs first, when the repair of the inoperable rod position indication can safely be performed. Actions 2.a), 2.b), and 2.c) shall not be allowed after the plant has been in MODE 5 or other plant condition, for a sufficient period of time, in which the repair of the inoperable rod position indication could have safely been performed. December 11, 2006 SEQUOYAH - UNIT 2 3/4 1-17 Amendment No. 235, 304 LCO 3.1.7 Applicability Add proposed ACTIONS Note 1 L01Add proposed ACTION DM01LA01ACTION B ACTION A Page 4 of 6 Required Action A.2 Note Add proposed ACTIONS Note 2A02 A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING 2.Place the control rods under manual control, and monitor and record Reactor CoolantSystem average temperature (Tavg) at least once per hour, and3.Restore the rod position indicators to OPERABLE status within 24 hours such that amaximum of one rod position indicator per bank is inoperable, or4.Be in HOT STANDBY within the next 6 hours.c.With a maximum of one demand position indicator per bank inoperable either:1.Verify that all rod position indicators for the affected bank are OPERABLE and that themost withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12steps of each other at least once per 12 hours, or2.Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8hours.SURVEILLANCE REQUIRMENTS 4.1.3.2 Each rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 12 steps at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours. December 11, 2006 SEQUOYAH - UNIT 2 3/4 1-17a Amendment No. 235, 304 Add proposed ACTION DM01M02Add proposed SR 3.1.7.1 ACTION B ACTION C ACTION D Page 5 of 6 A01ITS ITS 3.1.7 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 This specification is deleted. December 18, 2000 SEQUOYAH - UNIT 2 3/4 1-18 Amendment No. 15, 255 Page 6 of 6 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.1.3.2 Note

applies to Actions 2.a, 2.b, and 2.c and may be only applied to one inoperable rod position indicator. In this condition, the inoperable rod position indicator shall only be allowed until either the end of the current cycle, or until an entry into MODE 5 of sufficient duration, whichever occurs first, when the repair of the inoperable rod position indication can safely be performed. Actions 2.a, 2.b, and 2.c shall not be allowed after the plant has been in MODE 5 or other plant condition, for a sufficient period of time, in which the repair of the inoperable rod position indication could have safely been performed. ITS 3.1.7 ACTIONS Note 2 states that LCO 3.0.4.a and b are not applicable for Required Actions A.2.1 and A.2.2 following startup from a refueling outage, or following entry into MODE 5 of sufficient duration to safely repair an inoperable rod position indication. This changes the CTS by rewording the allowance for one rod position indicator inoperable to be consistent with ITS terminology. This change is designated as an administrative change since the change does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 3.1.3.2 ACTION a and c do not contain an ACTION to follow if the provided ACTIONS cannot be met. Therefore, CTS 3.0.3 would be entered, which would allow 1 hour to initiate a shutdown and 7 hours to be in HOT STANDBY.

ITS 3.1.7 ACTION D requires if the Required Actions and associated Completion Time of ACTION A or C are not met, to be in MODE 3 within 6 hours. This changes the CTS by eliminating the one hour to initiate a shutdown and consequently allows one hour less for the unit to be in MODE 3. This change is acceptable because it provides an appropriate compensatory measure for the described conditions. If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. The LCO is applicable in MODES 1 and 2. Requiring a shutdown to MODE 3 is appropriate in this condition. The one hour allowed by CTS 3.0.3 to prepare for a shutdown is not needed because the operators have had time to prepare for the shutdown while attempting to follow the Required Actions and associated Completion Times. This change is designated as more restrictive because it allows less time to shutdown than is allowed in the CTS. DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION Sequoyah Unit 1 and Unit 2 Page 2 of 3 M02 CTS 4.1.3.2 requires that each rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 12 steps at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours. ITS 3.1.7 does not contain this requirement because it is duplicative of CTS 4.1.3.1.1 (ITS SR 3.1.4.1). A new Surveillance has been added (ITS SR 3.1.7.1) to verify each RPI agrees within 12 steps of the group demand position for the full indicated range of rod travel, once prior to criticality after each removal of the reactor head. This changes the CTS by adding a new Surveillance Requirement. The purpose of ITS SR 3.1.7.1 is to provide additional assurance that the rod position indication system is operating correctly. This change is acceptable because it provides additional assurance that the rod position indication channels are OPERABLE. This change is designated as more restrictive because it adds a new Surveillance Requirement to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS LCO 3.1.3.2 requires the shutdown and control rod position indication system and the demand position indication system to be OPERABLE and capable of determining the control rod positions within +/- 12 steps. ITS LCO 3.1.7 requires the analog Rod Position Indication System and the Demand Position Indication System to be OPERABLE but the details of what constitutes an OPERABLE system are moved to the Bases. This changes the CTS by removing the details of what constitutes an OPERABLE system to the Bases. The removal of these details, which are related to system design, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement that the Rod Position Indication System and Demand Position Indication System be OPERABLE. The details on the capability requirements of the systems do not need to appear in the specification in order for the requirement to apply. Additionally, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION Sequoyah Unit 1 and Unit 2 Page 3 of 3 LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.3.2 ACTION a covers the inoperability for a maximum of one rod position indicator per bank. CTS 3.1.3.2 ACTION b covers the inoperability for more than one rod position indicator per bank. CTS 3.1.3.2 ACTION c covers the inoperability for a maximum of one demand position indicator per bank. ITS 3.1.7 ACTIONS are modified by Note 1 that states "Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator." ITS 3.1.7 ACTION A covers inoperability for one rod position indicator per bank. ITS 3.1.7 ACTION B covers inoperability for more than one rod position indicator per bank. ITS 3.1.7 ACTION C covers inoperability for one demand position indicator bank for one or more banks. This changes the CTS by allowing separate Condition entry for each inoperable rod position indicator and each demand position indicator. The purpose of CTS 3.1.3.2 ACTION a is to provide compensatory actions for a maximum of one rod position indicator per bank. The purpose of CTS 3.1.3.2 ACTION b is to provide compensatory actions for more than one rod position indicator per bank. The purpose of CTS 3.1.3.2 ACTION c is to provide compensatory actions for one demand position indicator per bank. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. This change will allow separate Condition entry for each inoperable rod position indicator and each inoperable demand position indicator while the CTS does not. The ITS will allow each inoperable rod position indicator or each inoperable demand position indicator to be tracked separately. This change is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for inoperable position indication. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. DOC L02 (Category 4 - Relaxation of Required Action) The CTS 3.1.3.2 ACTION for, "more than one rod position indicator per bank inoperable" requires the performance of ACTION b.1 and b.2 and b.3 or b.4. If CTS Actions b.1, b.2, and b.3 are not performed, then CTS 3.1.3.2 Action b.4 requires placing the unit in HOT STANDBY within the next 6 hours. CTS 3.1.3.2 Action b.1 requires, in part, determining the position of the non-indicating rod(s) indirectly by the movable incore detectors immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position. ITS 3.1.7 Condition C for, "one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position," requires performance of Required Action C.1 or C.2. ITS 3.1.7 Required Action C.1 requires verifying the position of the rods with inoperable position indicators indirectly by using the movable incore detectors with a Completion Time of immediately. ITS 3.1.7 Required Action C.2 requires reducing THERMAL POWER to < 50% RTP. This changes the CTS by allowing a reduction in THERMAL POWER as an alternative to verifying the position of the rods with inoperable position indicators and placing the unit in HOT STANDBY within the next 6 hours. The purpose of the CTS 3.1.3.2 ACTION to "verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors" is to ensure a misaligned rod does not go undetected and cause a power imbalance in the core. This change is acceptable because, if the rod positions have not been determined, THERMAL POWER must be reduced to < 50% RTP to avoid undesirable power distributions that could result from continued operation at 50% RTP when one or more rods are misaligned by more than 24 steps. This change is designated as less restrictive because less stringent Required Actions are being applied in ITS than were applied in CTS. Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-1Rev. 4.0 Amendment XXX SEQUOYAH UNIT 1 4CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The [Digital] Rod Position Indication ([D]RPI) System and the Demand Position Indication System shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS ------------------------------------------------------------NOTE----------------------------------------------------------- Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A. One [D]RPI per group inoperable for one or more groups. A.1 Verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors. OR A.2 Reduce THERMAL POWER to 50% RTP. Once per 8 hours 8 hours B. More than one [D]RPI per group inoperable. B.1 Place the control rods under manual control. AND B.2 Monitor and record Reactor Coolant System Tavg. AND Immediately Once per 1 hour INSERT 2INSERT 333.1.3.2 Applicability ACTION a ACTION b 1123314<12 4bank bank rod position indicator rod position indicator 4INSERT 1 1. S5 3.1.7 Insert Page 3.1.7-1a INSERT 1 2.LCO 3.0.4.a and b are not applicable for Required Actions A.2.1 and A.2.2 following astartup from a refueling outage, or following entry into MODE 5 of sufficient duration to safelyrepair an inoperable rod position indication.INSERT 2 AND Immediately after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position 243.1.3.2 Note* Action a.1 3.1.7 Insert Page 3.1.7-1b INSERT 3 OR --------------------NOTE------------------- Required Actions A.2.1 and A.2.2 may only be applied to one inoperable rod position indicator. ------------------------------------------------ A.2.1 Verify position of the rod with inoperable position indicator indirectly by using movable incore detectors. AND 8 hours AND Once per 31 days thereafter AND 8 hours if Rod Control System parameters indicate unintended movement AND 8 hours if the rod with an inoperable position indicator is moved greater than 12 steps AND Prior to increasing THERMAL POWER above 50% RTP AND 8 hours after reaching 100% RTP 3 3.1.7 Insert Page 3.1.7-1c INSERT 3 (Continued) A.2.2 Review the parameters of the Rod Control System for indications of unintended rod movement for the rod with the inoperable position indicator. 16 hours AND Once per 8 hours thereafter 3 Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-2Rev. 4.0 Amendment XXX SEQUOYAH UNIT 1 4CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 Verify the position of the rods with inoperable position indicators indirectly by using the movable incore detectors. AND B.4 Restore inoperable position indicators to OPERABLE status such that a maximum of one [D]RPI per group is inoperable. Once per 8 hours 24 hours C. One or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position. C.1 Verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors. OR C.2 Reduce THERMAL POWER to 50% RTP. [4] hours 8 hours D. One demand position indicator per bank inoperable for one or more banks. D.1.1 Verify by administrative means all [D]RPIs for the affected banks are OPERABLE. AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart. OR Once per 8 hours Once per 8 hours INSERT 4ACTION b ACTION c C C 22221112 412412 44rod position indicator rod position indicators bankDOC L01 4 3.1.7 Insert Page 3.1.7-2 INSERT 4 AND Immediately after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position 2Action b.1 Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-3Rev. 4.0 Amendment XXX SEQUOYAH UNIT 1 4CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D.2 Reduce THERMAL POWER to 50% RTP. 8 hours E. Required Action and associated Completion Time not met. E.1 Be in MODE 3. 6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify each [D]RPI agrees within [12] steps of the group demand position for the [full indicated range] of rod travel. Once prior to criticality after each removal of the reactor head ACTION c ACTION b.4, DOC M02 C D D 4.1.3.2 2214<at 20 and 215 stepsrod position indicator Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-1Rev. 4.0 Amendment XXX SEQUOYAH UNIT 2 4CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The [Digital] Rod Position Indication ([D]RPI) System and the Demand Position Indication System shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS ------------------------------------------------------------NOTE----------------------------------------------------------- Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A. One [D]RPI per group inoperable for one or more groups. A.1 Verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors. OR A.2 Reduce THERMAL POWER to 50% RTP. Once per 8 hours 8 hours B. More than one [D]RPI per group inoperable. B.1 Place the control rods under manual control. AND B.2 Monitor and record Reactor Coolant System Tavg. AND Immediately Once per 1 hour INSERT 2INSERT 333.1.3.2 Applicability ACTION a ACTION b 1123314<12 4bank bank rod position indicator rod position indicator 4INSERT 1 1. S5 3.1.7 Insert Page 3.1.7-1a INSERT 1 2.LCO 3.0.4.a and b are not applicable for Required Actions A.2.1 and A.2.2 following astartup from a refueling outage, or following entry into MODE 5 of sufficient duration to safelyrepair an inoperable rod position indication.INSERT 2 AND Immediately after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position 243.1.3.2 Note* Action a.1 3.1.7 Insert Page 3.1.7-1b INSERT 3 OR --------------------NOTE------------------- Required Actions A.2.1 and A.2.2 may only be applied to one inoperable rod position indicator. ------------------------------------------------ A.2.1 Verify position of the rod with inoperable position indicator indirectly by using movable incore detectors. AND 8 hours AND Once per 31 days thereafter AND 8 hours if Rod Control System parameters indicate unintended movement AND 8 hours if the rod with an inoperable position indicator is moved greater than 12 steps AND Prior to increasing THERMAL POWER above 50% RTP AND 8 hours after reaching 100% RTP 3 3.1.7 Insert Page 3.1.7-1c INSERT 3 (Continued) A.2.2 Review the parameters of the Rod Control System for indications of unintended rod movement for the rod with the inoperable position indicator. 16 hours AND Once per 8 hours thereafter 3 Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-2Rev. 4.0 Amendment XXX SEQUOYAH UNIT 2 4CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 Verify the position of the rods with inoperable position indicators indirectly by using the movable incore detectors. AND B.4 Restore inoperable position indicators to OPERABLE status such that a maximum of one [D]RPI per group is inoperable. Once per 8 hours 24 hours C. One or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position. C.1 Verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors. OR C.2 Reduce THERMAL POWER to 50% RTP. [4] hours 8 hours D. One demand position indicator per bank inoperable for one or more banks. D.1.1 Verify by administrative means all [D]RPIs for the affected banks are OPERABLE. AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart. OR Once per 8 hours Once per 8 hours INSERT 4ACTION b ACTION c C C 22221112 412412 44rod position indicator rod position indicators bankDOC L01 4 3.1.7 Insert Page 3.1.7-2 INSERT 4 AND Immediately after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position 2Action b.1 Rod Position Indication 3.1.7 Westinghouse STS 3.1.7-3Rev. 4.0 Amendment XXX SEQUOYAH UNIT 2 4CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D.2 Reduce THERMAL POWER to 50% RTP. 8 hours E. Required Action and associated Completion Time not met. E.1 Be in MODE 3. 6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify each [D]RPI agrees within [12] steps of the group demand position for the [full indicated range] of rod travel. Once prior to criticality after each removal of the reactor head ACTION c ACTION b.4, DOC M02 C D D 4.1.3.2 2214<at 20 and 215 stepsrod position indicator JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.The ISTS contains bracketed information and/or values that are generic toWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is inserted to reflect the current licensing basis.2.ISTS 3.1.7 ACTION C has been deleted and a new conditional Completion time hasbeen added to Required Action A.1 and B.3. The new completion time ensures thatSQN current licensing basis is maintained, in that a verification of the positionindicator is still being performed immediately after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position. Additionally, ISTS 3.1.7 ACTIONS D and E has been changed to ITS 3.1.7 ACTIONS C and D, respectively, because of this deletion.3.ISTS 3.1.7 ACTION A provides compensatory actions for when one rod positionindicator is inoperable. ITS 3.1.7 provides an additional Required Action that can be taken when one rod position indicator is inoperable. The new Required Action allows the use of an alternate means other than the movable incore detectors to monitor theposition of a control or shutdown rod when the analog rod position indication systemis inoperable. This change reflects a current licensing basis that was approved by the NRC in Amendment 315 for Unit 1 and Amendment 304 for Unit 2 (ADAMS Accession No. ML063120575). Additionally ISTS 3.1.7 Required Action A.2 has been renumbered as ITS 3.1.7 Required Action A.3.4.Changes are made (additions, deletions, and/or changes) to the ISTS that reflect theplant specific nomenclature, number, reference, system description, analysis, orlicensing basis description.5.Editorial changes made for enhanced clarity/consistency.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-1Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Rod Position Indication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE. LCO 3.1.7 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment. Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM. Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the [Digital] Rod Position Indication ([D]RPI) System. 2115resulting from5and shutdown Rod Position Indication B 3.1.7 Westinghouse STSB 3.1.7-2Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1BASES BACKGROUND (continued) The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. The [D]RPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the [D]RPI will go on half accuracy with an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the [D]RPI System is +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps (+/- 7.5 inches). With an indicated deviation of 12 steps between the group step counter and [D]RPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches. APPLICABLE Control and shutdown rod position accuracy is essential during power SAFETY operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions. The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident. 5/8125115Rod Position Indicationanare511INSERT 1 55 B 3.1.7 Insert Page B 3.1.7-2 INSERT 1 A deviation of +/- 12 steps between the group step counter and a rod position indication is based on normal Rod Position Indication System indication accuracy of +/- 5% span with a maximum uncertainty of 10% span between the group step counter and the rod position indication. 1 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-3Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1BASES LCO LCO 3.1.7 specifies that one [D]RPI System and one Bank Demand Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The [D]RPI System indicates within 12 steps of the group stepcounter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits," b. For the [D]RPI System there are no failed coils, andc. The Bank Demand Indication System has been calibrated either inthe fully inserted position or to the [D]RPI System. The 12 step agreement limit between the Bank Demand Position Indication System and the [D]RPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position. A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits). These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits. APPLICABILITY The requirements on the [D]RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),

because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System. 21Rod Position Indication 2211212Rod Position IndicationRod Position IndicationRod Position Indication Position4shall INSERT 2 INSERT 3Rod Position Indication5of5 B 3.1.7 Insert Page B 3.1.7-3 INSERT 2 Additionally, one Demand Position Indication System shall be OPERABLE for each group within a bank. INSERT 3 a check is performed between the two step counters in the same bank. Shutdown Banks C and D each contain a single group. Therefore, validation of movement for Shutdown Banks C and D can only be performed with a comparison of the single group to the corresponding RPI movement. 11 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-4Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1BASES ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator. A.1 When one [D]RPI channel per group fails, the position of the rod may still be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, HNF satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. A.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3). The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above. B.1, B.2, B.3, and B.4 When more than one [D]RPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via 23 44INSERT 5432<<44412Rod Position Indicationbank15Rod Position Indicationfails1bank5INSERT 44 B 3.1.7 Insert Page B 3.1.7-4 INSERT 4 A second Note has been added to provide clarification that LCO 3.0.4.a and LCO 3.0.4.c are not applicable for Required Action A.2.1 and A.2.2 following startup from a refueling outage, or following entry into MODE 5 of sufficient duration to safely repair an inoperable rod position indication. INSERT 5 If one or more rods have been significantly moved (in excess of 24 steps in one direction, since the position was last determined), Required Action A.1 is still appropriate, but actions must be initiated immediately to begin verifying that the rod is still properly positioned, relative to their group positions. In this Required Action, the Completion Time only begins on discovery that both: a.One rod position indication per bank is inoperable, andb.A rod with an inoperable position indicator has been moved in excess of 24 steps in onedirection since the last determination of the rod's position.If at any time during the existence of Condition A (one RPI per bank inoperable), a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position, this Completion Time begins to be tracked. A.2.1, and A.2.2 When one RPI channel per bank fails, the position of the rod may still be determined indirectly by use of the movable incore detectors and reviewing the parameters of the rod control system for indications of unintended rod movement for the rod with the inoperable position indication. Therefore, verification of RCCA position within 8 hours and every 31days thereafter is adequate for allowing continued full power operation as long as a review of the parameters of the rod control system for indications of unintended rod movement for the rod with the inoperable position indication is performed within 16 hours and every 8 hours thereafter. Furthermore, if the rod control system parameters indicate unintended movement or if the rod with an inoperable position indicator is moved greater than 12 steps, then the verification of the RCCA position must be performed within 8 hours. As long as these compensatory actions are met, reactor operation can then continue until the end of the current cycle or until an entry into MODE 5 of sufficient duration that the repair of the inoperable rod position indication can safely be performed. Required Actions A.2.1,and A.2.2 are modified by a Note directing that these Required Actions may only be applied to one inoperable rod position indicator. 44 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-5Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1BASES ACTIONS (continued) movable incore detectors will minimize the potential for rod misalignment. The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. Monitoring and recording reactor coolant Tavg help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions. The position of the rods may be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, HNF satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Verification of control rod position once per 8 hours is adequate for allowing continued full power operation for a limited, 24 hour period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour Completion Time provides sufficient time to troubleshoot and restore the [D]RPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication. Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required. C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2, [or B.1, as applicable] are still appropriate but must be initiated promptly under Required Action C.1 to begin verifying that these rods are still properly positioned, relative to their group positions. If, within [4] hours, the rod positions have not been determined, THERMAL POWER must be reduced to 50% RTP within 8 hours to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of [4] hours provides an acceptable period of time to verify the rod positions. INSERT 63244412Rod Position Indication helps 55 B 3.1.7 Insert Page B 3.1.7-5 INSERT 6 (in excess of 24 steps in one direction, since the position was last determined), Required Action B.3 is still appropriate, but action must be initiated immediately to begin verifying that the rod is properly positioned, relative to its bank position. In this Required Action, the Completion Time only begins on discovery that both: a.More than one RPI per bank is inoperable; andb.A rod with an inoperable position indicator has been moved in excess of 24 steps in onedirection since the last determination of the rod's position.If at any time during the existence of Condition B (more than one RPI per bank inoperable), a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position, this Completion Time begins to be tracked. 4 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-6Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX 1BASES ACTIONS (continued) D.1.1 and D.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the [D]RPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 12 steps apart within the allowed Completion Time of once every 8 hours is adequate. D.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours provides an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to 50% RTP. E.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the [D]RPI agrees with the demand position within [12] steps ensures that the [D]RPI is operating correctly. Since the [D]RPI does not display the actual shutdown rod positions between 18 and 210 steps, only points within the indicated ranges are required in comparison. This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power. C C D 44424241<<12 4Rod Position Indication Rod Position Indication1Rod Position Indication INSERT 7 B 3.1.7 Insert Page B 3.1.7-6 INSERT 7 This verification will be performed at 20 steps and 215 steps of rod travel. 1 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-7Rev. 4.0 Revision XXX SEQUOYAH UNIT 1 BASES REFERENCES 1.10 CFR 50, Appendix A, GDC 13.2. FSAR, Chapter [15].3. FSAR, Chapter[15].1212Section 7.7.1U Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-1Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Rod Position Indication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE. LCO 3.1.7 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment. Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM. Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the [Digital] Rod Position Indication ([D]RPI) System. 2115resulting from5and shutdown Rod Position Indication B 3.1.7 Westinghouse STSB 3.1.7-2Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1BASES BACKGROUND (continued) The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. The [D]RPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the [D]RPI will go on half accuracy with an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the [D]RPI System is +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps (+/- 7.5 inches). With an indicated deviation of 12 steps between the group step counter and [D]RPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches. APPLICABLE Control and shutdown rod position accuracy is essential during power SAFETY operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions. The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident. 5/8125115Rod Position Indicationanare511INSERT 1 55 B 3.1.7 Insert Page B 3.1.7-2 INSERT 1 A deviation of +/- 12 steps between the group step counter and a rod position indication is based on normal Rod Position Indication System indication accuracy of +/- 5% span with a maximum uncertainty of 10% span between the group step counter and the rod position indication. 1 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-3Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1BASES LCO LCO 3.1.7 specifies that one [D]RPI System and one Bank Demand Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The [D]RPI System indicates within 12 steps of the group stepcounter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits," b. For the [D]RPI System there are no failed coils, andc. The Bank Demand Indication System has been calibrated either inthe fully inserted position or to the [D]RPI System. The 12 step agreement limit between the Bank Demand Position Indication System and the [D]RPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position. A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits). These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits. APPLICABILITY The requirements on the [D]RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),

because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System. 21Rod Position Indication 2211212Rod Position IndicationRod Position IndicationRod Position Indication Position4shall INSERT 2 INSERT 3Rod Position Indication5of5 B 3.1.7 Insert Page B 3.1.7-3 INSERT 2 Additionally, one Demand Position Indication System shall be OPERABLE for each group within a bank. INSERT 3 a check is performed between the two step counters in the same bank. Shutdown Banks C and D each contain a single group. Therefore, validation of movement for Shutdown Banks C and D can only be performed with a comparison of the single group to the corresponding RPI movement. 11 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-4Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1BASES ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator. A.1 When one [D]RPI channel per group fails, the position of the rod may still be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, HNF satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. A.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3). The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above. B.1, B.2, B.3, and B.4 When more than one [D]RPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via 23 44INSERT 5432<<44412Rod Position Indicationbank15Rod Position Indicationfails1bank5INSERT 44 B 3.1.7 Insert Page B 3.1.7-4 INSERT 4 A second Note has been added to provide clarification that LCO 3.0.4.a and LCO 3.0.4.c are not applicable for Required Action A.2.1 and A.2.2 following startup from a refueling outage, or following entry into MODE 5 of sufficient duration to safely repair an inoperable rod position indication. INSERT 5 If one or more rods have been significantly moved (in excess of 24 steps in one direction, since the position was last determined), Required Action A.1 is still appropriate, but actions must be initiated immediately to begin verifying that the rod is still properly positioned, relative to their group positions. In this Required Action, the Completion Time only begins on discovery that both: a.One rod position indication per bank is inoperable, andb.A rod with an inoperable position indicator has been moved in excess of 24 steps in onedirection since the last determination of the rod's position.If at any time during the existence of Condition A (one RPI per bank inoperable), a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position, this Completion Time begins to be tracked. A.2.1, and A.2.2 When one RPI channel per bank fails, the position of the rod may still be determined indirectly by use of the movable incore detectors and reviewing the parameters of the rod control system for indications of unintended rod movement for the rod with the inoperable position indication. Therefore, verification of RCCA position within 8 hours and every 31days thereafter is adequate for allowing continued full power operation as long as a review of the parameters of the rod control system for indications of unintended rod movement for the rod with the inoperable position indication is performed within 16 hours and every 8 hours thereafter. Furthermore, if the rod control system parameters indicate unintended movement or if the rod with an inoperable position indicator is moved greater than 12 steps, then the verification of the RCCA position must be performed within 8 hours. As long as these compensatory actions are met, reactor operation can then continue until the end of the current cycle or until an entry into MODE 5 of sufficient duration that the repair of the inoperable rod position indication can safely be performed. Required Actions A.2.1,and A.2.2 are modified by a Note directing that these Required Actions may only be applied to one inoperable rod position indicator. 44 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-5Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1BASES ACTIONS (continued) movable incore detectors will minimize the potential for rod misalignment. The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. Monitoring and recording reactor coolant Tavg help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions. The position of the rods may be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, HNF satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Verification of control rod position once per 8 hours is adequate for allowing continued full power operation for a limited, 24 hour period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour Completion Time provides sufficient time to troubleshoot and restore the [D]RPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication. Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required. C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2, [or B.1, as applicable] are still appropriate but must be initiated promptly under Required Action C.1 to begin verifying that these rods are still properly positioned, relative to their group positions. If, within [4] hours, the rod positions have not been determined, THERMAL POWER must be reduced to 50% RTP within 8 hours to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of [4] hours provides an acceptable period of time to verify the rod positions. INSERT 63244412Rod Position Indication helps 55 B 3.1.7 Insert Page B 3.1.7-5 INSERT 6 (in excess of 24 steps in one direction, since the position was last determined), Required Action B.3 is still appropriate, but action must be initiated immediately to begin verifying that the rod is properly positioned, relative to its bank position. In this Required Action, the Completion Time only begins on discovery that both: a.More than one RPI per bank is inoperable; andb.A rod with an inoperable position indicator has been moved in excess of 24 steps in onedirection since the last determination of the rod's position.If at any time during the existence of Condition B (more than one RPI per bank inoperable), a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position, this Completion Time begins to be tracked. 4 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-6Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX 1BASES ACTIONS (continued) D.1.1 and D.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the [D]RPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 12 steps apart within the allowed Completion Time of once every 8 hours is adequate. D.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours provides an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to 50% RTP. E.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the [D]RPI agrees with the demand position within [12] steps ensures that the [D]RPI is operating correctly. Since the [D]RPI does not display the actual shutdown rod positions between 18 and 210 steps, only points within the indicated ranges are required in comparison. This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power. C C D 44424241<<12 4Rod Position Indication Rod Position Indication1Rod Position Indication INSERT 7 B 3.1.7 Insert Page B 3.1.7-6 INSERT 7 This verification will be performed at 20 steps and 215 steps of rod travel. 1 Rod Position Indication B 3.1.7 Westinghouse STS B 3.1.7-7Rev. 4.0 Revision XXX SEQUOYAH UNIT 2 BASES REFERENCES 1.10 CFR 50, Appendix A, GDC 13.2. FSAR, Chapter [15].3. FSAR, Chapter[15].1212Section 7.7.1U JUSTIFICATION FOR DEVIATIONS ITS 3.1.7 BASES, ROD POSITION INDICATION Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS Bases thatreflect the plant specific nomenclature, number, reference, system description,analysis, or licensing basis description.2.The ISTS contains bracketed information and/or values that are generic toWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is inserted to reflect the current licensing basis.3.ISTS 3.1.7 Required Action A.1 Bases contains a statement allowing an alternativemethod of satisfying Required Action A.1 by verifying that FQ and HNF are within thelimits provided in the COLR, provided the nonindicating rods have not been moved.Additionally, ISTS 3.1.7 Required Action B.3 Bases also contains this statement.ITS 3.1.7 Required Action A.1 Bases and Required Action B.3 Bases do not containthis statement. The statement has been deleted because it allows an alternative method for satisfying Required Actions A.1 and B.3 that are not addressed in the Specification. Since the Technical Specification Bases are not allowed to modify the Technical Specifications, this statement has been deleted.4.Changes are made to be consistent with changes made to the Specification.5.Editorial changes made for enhanced clarity/consistency. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.7, ROD POSITION INDICATION Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 8 ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.1.8 SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5 and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels low trip setpoints are set at less than or equal to 25% of RATED THERMAL POWER, and c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 531oF.

APPLICABILITY: MODE 2. ACTION: a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

b. With a Reactor Coolant System operating loop temperature (Tavg) less than 531oF, restore Tavg to within its limits within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 531oF at least once per 30 minutes during PHYSICS TESTS.

September 20, 2004 SEQUOYAH - UNIT 1 3/4 10-3 Amendment No. 295 Page 1 of 2 A03INSERT 13.1 REACTIVITY CONTROL SYSTEMS Exceptions - MODE 2 A02A04M01Add proposed LCO 3.1.8.b A05During PHYSICS TESTS initiated in Add proposed ACTION A M01M01LCO 3.1.8 ACTION B ACTION C SR 3.1.8.3 SR 3.1.8.1 SR 3.1.8.2 ACTION D Applicability OPERATIONAL M02LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program Add proposed SR 3.1.8.4 with a Frequency of 24 hours ITS 3.1.8 Insert Page 3/4 10-3 INSERT 1 and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 16.e, may be reduced to 3 required channels, A03 A01ITS ITS 3.1.8 SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at less than or equal to 25% of RATED THERMAL POWER, and

c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 531°F. APPLICABILITY: MODE 2. ACTION:

a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers. b. With a Reactor Coolant System operating loop temperature (Tavg) less than 531°F, restore (Tavg) to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 531°F at least once per 30 minutes during PHYSICS TESTS.

September 20, 2004 SEQUOYAH - UNIT 2 3/4 10-3 Amendment No. 285 Page 2 of 2 A03INSERT 13.1 REACTIVITY CONTROL SYSTEMS Exceptions - MODE 2 A02A04M01Add proposed LCO 3.1.8.b A05During PHYSICS TESTS initiated in Add proposed ACTION A M01Add proposed SR 3.1.8.4 with a Frequency of 24 hours M01LCO 3.1.8 SR 3.1.8.3 SR 3.1.8.1 SR 3.1.8.2 ACTION B ACTION C ACTION D Applicability LA01In accordance with the Surveillance Frequency Control Program LA01OPERATIONAL M02In accordance with the Surveillance Frequency Control Program ITS 3.1.8 Insert Page 3/4 10-3 INSERT 1 and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 16.e, may be reduced to 3 required channels, A03 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS Section 3.10 is titled SPECIAL TEST EXCEPTIONS. CTS Specification 3.10.3 is titled PHYSICS TESTS. ITS Section 3.1 is titled REACTIVITY CONTROL SYSTEMS. ITS Specification 3.1.8 is titled PHYSICS TESTS Exceptions - MODE 2. This changes the CTS by changing the title of the Section and the Specification. This change is acceptable because the requirements have not changed. This change is to the titles only. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 3.10.3 states the limitations of certain Specifications may be suspended during the performance of PHYSICS TESTS. ITS LCO 3.1.8 includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Function 2 (Power Range Neutron Flux), Function 3 (Power Range Neutron Flux Rate), Function 6, (Overtemperature T), and Function 16.e (Power Range Neutron Flux, P-10) from "4" to "3." This changes CTS 3.10.3 by adding an allowance to reduce the number of required RTS channels from "4" to "3" for specified Functions. The purpose of CTS 3.10.3 is to allow some flexibility during the performance of PHYSICS TESTS while ensuring appropriate limitations are in place to help ensure safe operation. This change is acceptable because the minimum channels required for OPERABILITY for these RTS Functions in CTS Table 3.3-1 is currently "3." This allowance is needed since the "Required Channels" in ITS 3.3.1, Reactor Trip System Instrumentation, is "4." The change from CTS "MINIMUM CHANNELS OPERABLE" to ITS "Required Channels is discussed in Discussion of Changes for ITS 3.3.1. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 3.10.3.b states that the limitations of certain Specifications may be suspended during the performance of PHYSICS TESTS provided the reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at less than or equal to 25% of RATED THERMAL POWER. ITS 3.1.8 states the requirements of certain Specifications may be suspended but contains no requirements on the Intermediate and Power Range Channels. The ITS contains the same requirements on the Intermediate and Power Range Channels in ITS LCO 3.3.1. This changes the CTS by eliminating the requirement that the Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 2 of 4 Channels are set at 25% of RATED THERMAL POWER from the test exception. This change is acceptable because the Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are contained in ITS LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." Repeating that requirement in the test exception LCO is unnecessary. This change is designated as administrative as it eliminates a repeated requirement from the CTS, resulting in no technical change to the CTS. A05 CTS 3.10.3 is applicable in MODE 2. ITS 3.1.8 is applicable during PHYSICS TESTS initiated in MODE 2. This changes the CTS such that the Specification is applicable in MODE 2 only when a PHYSICS TEST is initiated. The purpose of ITS 3.1.8 Applicability is to ensure the ACTIONS contained in the Specification are followed. The wording of the CTS appears to be contradictory because, if THERMAL POWER exceeds 5% RTP, then the test exception Specification Applicability is exited and the Actions no longer apply. However, it is clear that the CTS Action should be applied if THERMAL POWER exceeds 5% RTP and PHYSICS TESTS are in progress. The ITS Applicability eliminates this apparent contradiction and allows the test exception Conditions and Required Actions to be applied when the LCO is not met. This is consistent with the wording of the CTS ACTION. This change is designated as administrative because it clarifies the current wording of the Specification with no change in intent. MORE RESTRICTIVE CHANGES M01 CTS 3.10.3 states that limitations of certain Specifications may be suspended during the performance of PHYSICS TESTS and provides restrictions that must be followed when utilizing the CTS exception. ITS 3.1.8 adds a requirement that SHUTDOWN MARGIN must be within the limits provided in the COLR. A Surveillance (ITS SR 3.1.8.4), to verify the SHUTDOWN MARGIN every 24 hours, and an ACTION (ITS 3.1.8 ACTION A), to follow if the SHUTDOWN MARGIN is not met, are also added. See DOC LA01 for the discussion on moving the 24 hours Frequency to the Surveillance Frequency Control Program. This changes the CTS by imposing an additional requirement on the application of the test exception LCO. This change is acceptable because it imposes reasonable restrictions on the performance of PHYSICS TESTS when the control rod and RCS minimum temperature Specifications are allowed to be violated. The Bases for ITS 3.1.1, "SHUTDOWN MARGIN," states that during MODE 2, the SHUTDOWN MARGIN is ensured by compliance with the rod insertion limit Specifications. Under this test exception, those limits are allowed to be violated. This change is designated as more restrictive because it imposes additional restrictions not found in the CTS. M02 CTS 4.10.3.2 requires performance of a CHANNEL FUNCTIONAL TEST on each Intermediate and Power Range Channel. ITS SR 3.1.8.1 requires DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 3 of 4 performance of a CHANNEL OPERATIONAL TEST (COT) on each intermediate and power range channel. This changes the CTS by requiring a COT instead of a CHANNEL FUNCTIONAL TEST. CTS defines a CHANNEL FUNCTIONAL TEST as the injection of a simulated signal into the sensor as close to the sensor as practicable to verify OPERABILITY. ITS defines a COT as the injection of an actual or simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. This changes the CTS by requiring adjustments of the setpoints so that the Intermediate and Power Range Channel are within the necessary range and accuracy. This change is designated as more restrictive because it imposes additional requirements on testing.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.10.3.1 requires determining that the THERMAL POWER is less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. CTS 4.10.3.3 requires determining that the Reactor Coolant System temperature (Tavg) is greater than or equal to 531°F at least once per 30 minutes during PHYSICS TESTS. ITS SR 3.1.8.2 and ITS SR 3.1.8.3 requires similar Surveillances and specifies the periodic Frequencies as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for these SR and associated Bases to the Surveillance Frequency Control Program.

The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 4 of 4 as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-1 Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions - MODE 2

LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of: LCO 3.1.3, "Moderator Temperature Coefficient," LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, "Control Bank Insertion Limits," and LCO 3.4.2, "RCS Minimum Temperature for Criticality"

may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e, may be reduced to 3 required channels, provided: a. RCS lowest loop average temperature is [531]°F, b. SDM is within the limits specified in the COLR, and c. THERMAL POWER is < 5% RTP.

APPLICABILITY: During PHYSICS TESTS initiated in MODE 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore SDM to within limit. AND A.2 Suspend PHYSICS TESTS exceptions. 15 minutes

1 hour B. THERMAL POWER not within limit. B.1 Open reactor trip breakers. Immediately C. RCS lowest loop average temperature not within limit. C.1 Restore RCS lowest loop average temperature to within limit. 15 minutes 3.10.3 Applicability DOC M01 ACTION a ACTION b 162;;4 PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-2 Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition C not met. D.1 Be in MODE 3. 15 minutes

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on power range and intermediate range channels per [SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1]. Prior to initiation of PHYSICS TESTS SR 3.1.8.2 Verify the RCS lowest loop average temperature is [531]°F. [ 30 minutes OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. [ 30 minutes OR In accordance with the Surveillance Frequency Control Program ] ACTION b 4.10.3.2 4.10.3.3 4.10.3.1 333311 PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-3 Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.8.4 Verify SDM is within the limits specified in the COLR. [ 24 hours OR In accordance with the Surveillance Frequency Control Program ] DOC M01 33 PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-1 Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions - MODE 2

LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of: LCO 3.1.3, "Moderator Temperature Coefficient," LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, "Control Bank Insertion Limits," and LCO 3.4.2, "RCS Minimum Temperature for Criticality"

may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e, may be reduced to 3 required channels, provided: a. RCS lowest loop average temperature is [531]°F, b. SDM is within the limits specified in the COLR, and c. THERMAL POWER is < 5% RTP.

APPLICABILITY: During PHYSICS TESTS initiated in MODE 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore SDM to within limit. AND A.2 Suspend PHYSICS TESTS exceptions. 15 minutes

1 hour B. THERMAL POWER not within limit. B.1 Open reactor trip breakers. Immediately C. RCS lowest loop average temperature not within limit. C.1 Restore RCS lowest loop average temperature to within limit. 15 minutes 3.10.3 Applicability DOC M01 ACTION a ACTION b 162;;4 PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-2 Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition C not met. D.1 Be in MODE 3. 15 minutes

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on power range and intermediate range channels per [SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1]. Prior to initiation of PHYSICS TESTS SR 3.1.8.2 Verify the RCS lowest loop average temperature is [531]°F. [ 30 minutes OR In accordance with the Surveillance Frequency Control Program ] SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. [ 30 minutes OR In accordance with the Surveillance Frequency Control Program ] ACTION b 4.10.3.2 4.10.3.3 4.10.3.1 333311 PHYSICS TESTS Exceptions - MODE 2 3.1.8 Westinghouse STS 3.1.8-3 Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.8.4 Verify SDM is within the limits specified in the COLR. [ 24 hours OR In accordance with the Surveillance Frequency Control Program ] DOC M01 33 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8, PHYSICS TEST EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS SR 3.1.8.2, SR 3.1.8.3, and SR 3.1.8.4 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. 4. The punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-1 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.8 PHYSICS TESTS Exceptions - MODE 2

BASES BACKGROUND The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to be performed. Section XI of 10 CFR 50, Appendix B (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that the specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested. This testing is an integral part of the design, construction, and operation of the plant. Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59 (Ref. 2).

The key objectives of a test program are to (Ref. 3):

a. Ensure that the facility has been adequately designed,

b. Validate the analytical models used in the design and analysis,

c. Verify the assumptions used to predict unit response,

d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design, and

e. Verify that the operating and emergency procedures are adequate.

To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4).

PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of the testing required to ensure that the design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation. 5The5the PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-2 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

BACKGROUND (continued) The PHYSICS TESTS required for reload fuel cycles (Ref. 4) in MODE 2 are listed below: a. Critical Boron Concentration - Control Rods Withdrawn,

b. Critical Boron Concentration - Control Rods Inserted,

c. Control Rod Worth,

d. Isothermal Temperature Coefficient (ITC), and e. Neutron Flux Symmetry. The first four tests are performed in MODE 2, and the last test can be performed in either MODE 1 or 2. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.

  [ a. The Critical Boron Concentration - Control Rods Withdrawn Test measures the critical boron concentration at hot zero power (HZP). With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs. b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having a worth of at least 1% k/k when fully inserted into the core. This test is used to measure the boron reactivity coefficient. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The selected bank is then inserted to make up for the decreasing boron concentration until the selected bank has been moved over its entire range of travel. The reactivity resulting from each incremental bank movement is measured with a reactivity computer. The difference between the measured critical boron concentration with all rods fully withdrawn and with the bank inserted is determined. The boron reactivity coefficient is determined by dividing the measured bank worth by the measured boron concentration difference. Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits,"

LCO 3.1.5, "Shutdown Bank Insertion Limit," or LCO 3.1.6, "Control Bank Insertion Limits." 2. ; 1; and6111 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-3 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

BACKGROUND (continued) c. The Control Rod Worth Test is used to measure the reactivity worth of selected control banks. This test is performed at HZP and has three alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected control bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining control banks. The second method, the Rod Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn. The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining control banks. The third method, the Boron Endpoint Method, moves the selected control bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected control bank. This sequence is repeated for the remaining control banks. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."

e. The Flux Symmetry Test measures the degree of azimuthal symmetry of the neutron flux at as low a power level as practical, depending on the test method employed. This test can be performed at HZP (Control Rod Worth Symmetry Method) or at 30% RTP (Flux Distribution Method). The Control Rod Worth Symmetry 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-4 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

BACKGROUND (continued)

Method inserts a control bank, which can then be withdrawn to compensate for the insertion of a single control rod from a symmetric set. The symmetric rods of each set are then tested to evaluate the symmetry of the control rod worth and neutron flux (power distribution). A reactivity computer is used to measure the control rod worths. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. The Flux Distribution Method uses the incore flux detectors to measure the azimuthal flux distribution at selected locations with the core at 30% RTP. ] APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for development ANALYSES of the LCOs that are excepted by this LCO are described in the Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5). The above mentioned PHYSICS TESTS, and other tests that may be required to calibrate nuclear instrumentation or to diagnose operational problems, may require the operating control or process variables to deviate from their LCO limitations. The FSAR defines requirements for initial testing of the facility, including PHYSICS TESTS. Tables [14.1-1 and 14.1-2] summarize the zero, low power, and power tests. Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to 5% RTP, the reactor coolant temperature is kept 531°F, and SDM is within the limits provided in the COLR. The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. U 2121Core Operating Limit Methodology for Westinghouse Designed PWRs 15the5representing that11997 s 5 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-5 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES APPLICABLE SAFETY ANALYSES (continued) As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity. LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One power range neutron flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met. The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e may be reduced to 3 required channels during the performance of PHYSICS TESTS provided: a. RCS lowest loop average temperature is [531]°F, b. SDM is within the limits provided in the COLR, and

c. THERMAL POWER is 5% RTP. APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS. The Applicability is stated as "during PHYSICS TESTS initiated in MODE 2" to ensure that the 5% RTP maximum power level is not exceeded. Should the THERMAL POWER exceed 5% RTP, and consequently the unit enter MODE 1, this Applicability statement prevents exiting this Specification and its Required Actions. 126 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-6 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

ACTIONS A.1 and A.2 If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit. Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.

B.1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits. C.1 When the RCS lowest Tavg is < 531°F, the appropriate action is to restore Tavg to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 531°F could violate the assumptions for accidents analyzed in the safety analyses. D.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems. PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-7 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.

SR 3.1.8.2 Verification that the RCS lowest loop Tavg is 531°F will ensure that the unit is not operating in a condition that could invalidate the safety analyses. [ Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

-----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.8.3 Verification that the THERMAL POWER is 5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. [ Verification of the THERMAL POWER at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 3433 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-8 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES SURVEILLANCE REQUIREMENTS (continued) -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.8.4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration, b. Control bank position, c. RCS average temperature, d. Fuel burnup based on gross thermal energy generation,

e. Xenon concentration,

f. Samarium concentration,

g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH), h. Moderate defect, when above the POAH, and i. Doppler defect, when above the POAH.

Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.

[ The Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 433Moderator temperature 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-9 Rev. 4.0 1SEQUOYAH UNIT 1 Revision XXX BASES SURVEILLANCE REQUIREMENTS (continued) -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August, 1978. 4. ANSI/ANS-19.6.1-1985, December 13, 1985.

5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.

6. WCAP-11618, including Addendum 1, April 1989. 431BAW-10163P-A, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989 119971 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-1 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX B 3.1 REACTIVITY CONTROL SYSTEMS

B 3.1.8 PHYSICS TESTS Exceptions - MODE 2

BASES BACKGROUND The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to be performed. Section XI of 10 CFR 50, Appendix B (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that the specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested. This testing is an integral part of the design, construction, and operation of the plant. Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59 (Ref. 2).

The key objectives of a test program are to (Ref. 3):

a. Ensure that the facility has been adequately designed,

b. Validate the analytical models used in the design and analysis,

c. Verify the assumptions used to predict unit response,

d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design, and

e. Verify that the operating and emergency procedures are adequate.

To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4).

PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of the testing required to ensure that the design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation. 5The5the PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-2 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

BACKGROUND (continued) The PHYSICS TESTS required for reload fuel cycles (Ref. 4) in MODE 2 are listed below: a. Critical Boron Concentration - Control Rods Withdrawn,

b. Critical Boron Concentration - Control Rods Inserted,

c. Control Rod Worth,

d. Isothermal Temperature Coefficient (ITC), and e. Neutron Flux Symmetry. The first four tests are performed in MODE 2, and the last test can be performed in either MODE 1 or 2. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.

  [ a. The Critical Boron Concentration - Control Rods Withdrawn Test measures the critical boron concentration at hot zero power (HZP). With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs. b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having a worth of at least 1% k/k when fully inserted into the core. This test is used to measure the boron reactivity coefficient. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The selected bank is then inserted to make up for the decreasing boron concentration until the selected bank has been moved over its entire range of travel. The reactivity resulting from each incremental bank movement is measured with a reactivity computer. The difference between the measured critical boron concentration with all rods fully withdrawn and with the bank inserted is determined. The boron reactivity coefficient is determined by dividing the measured bank worth by the measured boron concentration difference. Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits,"

LCO 3.1.5, "Shutdown Bank Insertion Limit," or LCO 3.1.6, "Control Bank Insertion Limits." 2. ; 1; and6111 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-3 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

BACKGROUND (continued) c. The Control Rod Worth Test is used to measure the reactivity worth of selected control banks. This test is performed at HZP and has three alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected control bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining control banks. The second method, the Rod Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn. The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining control banks. The third method, the Boron Endpoint Method, moves the selected control bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected control bank. This sequence is repeated for the remaining control banks. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."

e. The Flux Symmetry Test measures the degree of azimuthal symmetry of the neutron flux at as low a power level as practical, depending on the test method employed. This test can be performed at HZP (Control Rod Worth Symmetry Method) or at 30% RTP (Flux Distribution Method). The Control Rod Worth Symmetry 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-4 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

BACKGROUND (continued)

Method inserts a control bank, which can then be withdrawn to compensate for the insertion of a single control rod from a symmetric set. The symmetric rods of each set are then tested to evaluate the symmetry of the control rod worth and neutron flux (power distribution). A reactivity computer is used to measure the control rod worths. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. The Flux Distribution Method uses the incore flux detectors to measure the azimuthal flux distribution at selected locations with the core at 30% RTP. ] APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for development ANALYSES of the LCOs that are excepted by this LCO are described in the Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5). The above mentioned PHYSICS TESTS, and other tests that may be required to calibrate nuclear instrumentation or to diagnose operational problems, may require the operating control or process variables to deviate from their LCO limitations. The FSAR defines requirements for initial testing of the facility, including PHYSICS TESTS. Tables [14.1-1 and 14.1-2] summarize the zero, low power, and power tests. Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to 5% RTP, the reactor coolant temperature is kept 531°F, and SDM is within the limits provided in the COLR. The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. U 2121Core Operating Limit Methodology for Westinghouse Designed PWRs 15the5representing that11997 3s 5 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-5 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES APPLICABLE SAFETY ANALYSES (continued) As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity. LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One power range neutron flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met. The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e may be reduced to 3 required channels during the performance of PHYSICS TESTS provided: a. RCS lowest loop average temperature is [531]°F, b. SDM is within the limits provided in the COLR, and

c. THERMAL POWER is 5% RTP. APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS. The Applicability is stated as "during PHYSICS TESTS initiated in MODE 2" to ensure that the 5% RTP maximum power level is not exceeded. Should the THERMAL POWER exceed 5% RTP, and consequently the unit enter MODE 1, this Applicability statement prevents exiting this Specification and its Required Actions. 126 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-6 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

ACTIONS A.1 and A.2 If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit. Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.

B.1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits. C.1 When the RCS lowest Tavg is < 531°F, the appropriate action is to restore Tavg to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 531°F could violate the assumptions for accidents analyzed in the safety analyses. D.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems. PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-7 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.

SR 3.1.8.2 Verification that the RCS lowest loop Tavg is 531°F will ensure that the unit is not operating in a condition that could invalidate the safety analyses. [ Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

-----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.8.3 Verification that the THERMAL POWER is 5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. [ Verification of the THERMAL POWER at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 3433 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-8 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES SURVEILLANCE REQUIREMENTS (continued) -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.1.8.4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration, b. Control bank position, c. RCS average temperature, d. Fuel burnup based on gross thermal energy generation,

e. Xenon concentration,

f. Samarium concentration,

g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH), h. Moderate defect, when above the POAH, and i. Doppler defect, when above the POAH.

Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.

[ The Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 433Moderator temperature 1 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 Westinghouse STS B 3.1.8-9 Rev. 4.0 1SEQUOYAH UNIT 2 Revision XXX BASES SURVEILLANCE REQUIREMENTS (continued) -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August, 1978. 4. ANSI/ANS-19.6.1-1985, December 13, 1985.

5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.

6. WCAP-11618, including Addendum 1, April 1989. 431BAW-10163P-A, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989 119971 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8 BASES, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. ISTS SR 3.1.8.2, SR 3.1.8.3, and SR 3.1.8.4 Bases provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 5. Editorial changes made for enhanced clarity/consistency. 6. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 9 Relocated/Deleted Current Technical Specifications (CTS) CTS 3/4.10.1, SHUTDOWN MARGIN Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) CTS 3/4.10.13/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s). APPLICABILITY: MODE 2. ACTION: a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

November 26, 1993 SEQUOYAH - UNIT 1 3/4 10-1 Amendment No. 12, 172 M01 CTS 3/4.10.13/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s). APPLICABILITY: MODE 2. ACTION: a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 35 gpm of a solution containing greater than or equal to 6120 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

November 26, 1993 SEQUOYAH - UNIT 2 3/4 10-1 Amendment No. 163 M01 DISCUSSION OF CHANGES CTS 3/4.10.1, SHUTDOWN MARGIN Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES None

MORE RESTRICTIVE CHANGES M01 CTS 3.10.1 provides an exception to the SHUTDOWN MARGIN requirements in CTS 3.1.1.1 in MODE 2 due to the purpose of the measurement of rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s). According to the Bases, this special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations. This changes the CTS by eliminating a special test exception. This change is acceptable because this method of testing is no longer used. As a result, the CTS special test exception is not needed. Other rod worth measurement techniques that do not violate the SHUTDOWN MARGIN requirements are used. This change is designated as more restrictive because an exception to the CTS is being deleted.

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES None

LESS RESTRICTIVE CHANGES None

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.10.1, SHUTDOWN MARGIN Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. CTS 3/4.10.2, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) CTS 3/4.10.2SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1 ACTION: With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specification 3.13.1., 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 Perform the surveillance required by the below listed Specifications at least once per 12 hours during PHYSICS TESTS:

a. Specification 4.2.2.2 and 4.2.2.3

b. Specification 4.2.3.2.

SEQUOYAH - UNIT 1 3/4 10-2 September 17, 1980 M01 CTS 3/4.10.2SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1. ACTION: With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 Perform the surveillance required by the below listed Specifications at least once per 12 hours during PHYSICS TESTS:

a. Specification 4.2.2 2 and 4.2.2.3 b. Specification 4.2.3.2

SEQUOYAH - UNIT 2 3/4 10-2 M01 DISCUSSION OF CHANGES CTS 3/4.10.2, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES None

MORE RESTRICTIVE CHANGES M01 CTS 3/4.10.2 provides an exception to the rod group height, rod insertion, and power distribution limits specifications. This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure control rod worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions. The ITS does not contain this special test exception. This changes the CTS by eliminating a special test exception. This change is acceptable because these types of PHYSICS TESTS (measurement of control rod worth and determination of the reactor stability index as well as the damping factor under xenon oscillation conditions) are only performed during initial plant startup test programs. These tests are not performed during post-refueling PHYSICS TESTS. As a result, the CTS special test exception is not needed. This change is designated as more restrictive because an exception to the CTS is being deleted. RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.10.2, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. CTS 3/4.10.4, REACTOR COOLANT LOOPS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) CTS 3/4.10.1SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set less than or equal to 25% of RATED THERMAL POWER APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS.

4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating startup or PHYSICS TESTS.

September 20, 2004 SEQUOYAH - UNIT 1 3/4 10-4 Amendment No. 295 M01 CTS 3/4.10.1SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN 3.10.5 This specification is deleted.

December 18, 2000 SEQUOYAH - UNIT 1 3/4 10-5 Amendment No. 1, 264 A01 CTS 3/4.10.4SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of start up and PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during start up and PHYSICS TESTS. 4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating start up and PHYSICS TESTS.

September 20, 2004 SEQUOYAH - UNIT 2 3/4 10-4 Amendment No. 285 M01 CTS 3/4.10.4SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.5 This specification is deleted.

December 18, 2000 SEQUOYAH - UNIT 2 3/4 10-5 Amendment No. 255 A01 DISCUSSION OF CHANGES CTS 3/4.10.4, REACTOR COOLANT LOOPS Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3/4.10.4 provides an exception to the reactor coolant loops Specification. This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. Testing within the required frequency is sufficient for verification that the power range and intermediate range monitors are properly functioning. The ITS does not contain this special test exception. This changes the CTS by eliminating a special test exception. This change is acceptable because these types of PHYSICS TESTS are no longer performed. Future PHYSICS TESTS will be performed under 3.1.8, "PHYSICS TESTS Exceptions - MODE 2." As a result this CTS Special test exception is not needed. This change is designated as more restrictive because an exception to the CTS is being deleted.

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES None

LESS RESTRICTIVE CHANGES None

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.10.4, REACTOR COOLANT LOOPS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ENCLOSURE 2 VOLUME 7 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.2 POWER DISTRIBUTION LIMITS Revision 0 LIST OF ATTACHMENTS 1. ITS 3.2.1, - Heat Flux Hot Channel Factor (FQ(X,Y, Z)) 2. ITS 3.2.2, - Nuclear Enthalpy Rise Hot Channel Factor (FH(X, Y)) 3. ITS 3.2.3 - Axial Flux Difference (AFD) 4. ITS 3.2.4 - Quadrant Power Tilt Ratio (QPTR) ATTACHMENT 1 ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-FQ(X,Y,Z) LIMITING CONDITION FOR OPERATION 3.2.2 FQ(X,Y,Z) shall be maintained within the acceptable limits specified in the COLR: APPLICABILITY: MODE 1 ACTION: With FQ(X,Y,Z) exceeding its limit: a. Reduce THERMAL POWER at least 1% for each 1% FQ(X,Y,Z) exceeds the limit within 15 minutes, and similarly reduce the following:

1. Administratively reduce the allowable power at each point along the AFD limit lines within 2 hours, and

2. The Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. b. POWER OPERATION may proceed for up to 48 hours. Subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K4) have been reduced at least 1% (in T span) for each 1% that FQ(X,Y,Z) exceeds the limit specified in the COLR.

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Action a. and b., above; THERMAL POWER may then be increased provided FQ(X,Y,Z) is demonstrated through incore mapping to be within its limits.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-5 Amendment No. 19, 95, 140, 155, 223 Page 1 of 10 LCO 3.2.1 Applicability ACTION A Required Action A.4 Add proposed ACTION A NoteM01Required Action A.1 Required Action A.3 Required Action A.5 LA01LA02Add proposed ACTION D M03Required Action A.2 SR NOTE M04M02after each FQ(X,Y,Z) determination )Z,Y,X(FCQsteady state A02A02)Z,Y,X(FCQL0172 M02after each FQ(X,Y,Z) determination A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 QMF(X,Y,Z) shall be evaluated to determine if FQ(X,Y,Z) is within its limit by: a. Using the moveable incore detectors to obtain a power distribution map (QMF(X,Y,Z)*) at any THERMAL POWER greater than 5% of RATED THERMAL POWER. b. Satisfying the following relationship: (QMF(X,Y,Z) BQNOM(X,Y,Z) where BQNOM (X,Y,Z)** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement. The BQNOM (X,Y,Z) factors are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive. c. If the above relationship is not satisfied, then 1. For that location, calculate the % margin to the maximum allowable design as follows: where BQDES(X,Y,Z)** and BCDES(X,Y,Z)** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting Condition for Operation limits, and include allowances for the calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for QMF(X,Y,Z), because the limits include uncertainties.
- BQNOM (X,Y,Z), BQDES(X,Y,Z), and BCDES(X,Y,Z) Data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-6 Amendment Nos. 19, 95, 140, 155, 223 % AFD Margin = 1 - F(X,Y,Z)BQDES(X,Y,Z) x 100%QM % f(I) Margin = 1 - F(X,Y,Z)BCDES(X,Y,Z) x 100%2QM SR 3.2.1.2 SR 3.2.1.3 Page 2 of 10 LA03SR 3.2.1.2 SR 3.2.1.3 LA03LA03LA03 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above. AFD min margin = minimum % margin value of all locations examined. f2(I) OPT min margin = minimum % margin value of all locations examined.

3. If the AFD min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

  (a) Within 2 hours, administratively reduce the negative AFD limit lines at each power level by:

Reduced AFDLimit = (AFDLimit from COLR) + absolute value of (NSLOPEAFD* % x AFD min margin of 4.2.2.2.c.2)

  (b) Within 2 hours, administratively reduce the positive AFD limit lines at each power level by:

Reduced AFDLimit = (AFDLimit from COLR) absolute value of (PSLOPEAFD* % X AFD min margin)

4. If the f2(I) min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

  (a) Within 48 hours, reduce the OPT negative f2(I) breakpoint limit by:

Reduced OPT negative f2(I) breakpoint limit = (f2(I) limit of Table 2.2-1) + absolute value of

NSLOPEAFD and PSLOPEAFD are the amount of AFD adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 ** PSLOPE and NSLOPEI)(fI)(f22 are the amounts of the OPT f2(I) limit adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-7 Amendment No. 19, 95, 140, 155, 216, 223 margin) min I)(f x % NSLOPE(2)I(f**2 LA03Page 3 of 10 ACTION B REQUIRED ACTION B.2 REQUIRED ACTION B.1 REQUIRED ACTION C.2 ACTION C REQUIRED ACTION B.1/B.2 REQUIRED ACTION C.1/C.2 LA03LA03LA03LA03M05M05LA03 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) (b) Within 48 hours, reduce the OPT positive f2(I) breakpoint limit by: Reduced OPT positive f2(I) breakpoint limit = (f2(I) limit of Table 2.2-1) absolute value of margin) min I)(f x % **PSLOPE(2I)(f2 d. Measuring QMF(X,Y,Z) according to the following schedule: 1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(X,Y,Z) was last determined,*** or 2. At least once per 31 Effective Full Power Days, whichever occurs first.

e. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding QMF(X,Y,Z) BQNOM(X,Y,Z), either of the following actions specified shall be taken. 1. QMF(X,Y,Z) shall be increased over that specified in 4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or 2. QMF(X,Y,Z) shall be evaluated according to 4.2.2.2 at or before the time when the margin is projected to result in one of the actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4. 4.2.2.3 When FQ(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(X,Y,Z) shall be obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the FQ(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

___________________________ ** PSLOPE and NSLOPEI)(fI)(f22 are the amounts of the OPT f2(I) limit adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 *** During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-8 Amendment No. 19, 140, 223 Page 4 of 10 M08thereafter LA04In accordance with the Surveillance Frequency Control ProgramSR Note INSERT 1 andM06SR 3.2.1.1 SR 3.2.1.2 SR 3.2.1.3 LA02LA03REQUIRED ACTION C.1 SR 3.2.1.2/SR 3.2.1.3 NOTE SR 3.2.1.2 / SR 3.2.1.3 NOTE a. SR 3.2.1.2 / SR 3.2.1.3 NOTE b. SR 3.2.1.1 LA03REQUIRED ACTION C.1/C.2 M08M04M07M09INSERT 2 Once within 12 hours afterA02)Z,Y,X(FCQNot required to be performed until 12 hours after an equilibrium can be 3.2.1 Insert Page 3/4 2-8 CTS (Page 4 of 10) INSERT 1 Once after each refueling prior to THERMAL POWER exceeding 75% RTP

INSERT 2 Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FCQ was last verified AND At least once per 31 Effective Full Power Days DOC M09 DOC M06 LA04In accordance with the Surveillance Frequency Control Program A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS

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October 23, 1991 SEQUOYAH - UNIT 1 3/4 2-9 Amendment No. 12, 140, 155 Page 5 of 10 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR- FQ(X,Y,Z) LIMITING CONDITION FOR OPERATION 3.2.2 FQ(X,Y,Z) shall be maintained within the acceptable limits specified in the COLR. APPLICABILITY: MODE 1 ACTION: With FQ(X,Y,Z) exceeding its limit: a. Reduce THERMAL POWER at least 1% for each 1% FQ(X,Y,Z) exceeds the limit within 15 minutes, and similarly reduce the following: 1. Administratively reduce the allowable power at each point along the AFD limit lines within 2 hours, and

2. The Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.

b. POWER OPERATION may proceed for up to 48 hours. Subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K4) have been reduced at least 1% (in T span) for each 1% that FQ(X,Y,Z) exceeds the limit specified in the COLR. c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Action a. and b., above; THERMAL POWER may then be increased provided FQ(X,Y,Z) is demonstrated through incore mapping to be within its limits.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-4 Amendment Nos. 21, 95, 131, 146, 214 LCO 3.2.1 Applicability ACTION A Required Action A.4 Required Action A.1 Required Action A.3 Required Action A.5 LA01Page 6 of 10 M01Add proposed ACTION DM03Required Action A.2 SR NOTE M04M02after each FQ(X,Y,Z) determination A02)Z,Y,X(FCQAdd proposed ACTION A Notesteady state A02)Z,Y,X(FCQLA02M02after each FQ(X,Y,Z) determination L0172 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 Z)Y,(X,FMQ shall be evaluated to determine if FQ(X,Y,Z) is within its limit by: a. Using the moveable incore detectors to obtain a power distribution map (QMF(X,Y,Z)*) at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Satisfying the following relationship:

QMF(X,Y,Z) BQNOM(X,Y,Z) where BQNOM (X,Y,Z)** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement. The BQNOM (X,Y,Z) factors are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive. 2. Upper core region from 85 to 100%, inclusive.

c. If the above relationship is not satisfied, then 1. For that location, calculate the % margin to the maximum allowable design as follows: where BQDES (X,Y,Z)** and BCDES(X,Y,Z)** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting Condition for Operation limits, and include allowances for the calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for QMF(X,Y,Z), because the limits include uncertainties. ** BQNOM (X,Y,Z), BQDES (X,Y,Z), and BCDES (X,Y,Z) Data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-5 Amendment No. 21, 95, 131, 146, 214 % AFD Margin = 1 - F(X,Y,Z)BQDES(X,Y,Z) x 100%QM % f(I) Margin = 1 - F(X,Y,Z)BCDES(X,Y,Z) x 100%2QMPage 7 of 10 LA03SR 3.2.1.2 SR 3.2.1.3 LA03SR 3.2.1.2 SR 3.2.1.3 LA03 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above. AFD min margin = minimum % margin value of all locations examined. f2(I) OPT min margin = minimum % margin value of all locations examined.

3. If the AFD min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

  (a) Within 2 hours, administratively reduce the negative AFD limit lines at each power level by:           Reduced AFDLimit = (AFDLimit from COLR) + absolute value of (NSLOPEAFD* % x AFD min margin of 4.2.2.2.c.2)       
  (b) Within 2 hours, administratively reduce the positive AFD limit lines at each power level by:

Reduced AFDLimit = (AFDLimit from COLR)-absolute value of (PSLOPEAFD* % X AFD min margin)

4. If the f2(I) min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed. (a) Within 48 hours, reduce the OPT negative f2(I) breakpoint limit by:

Reduced OPT negative f2(I) breakpoint limit = (f2(I) limit of Table 2.2-1) + absolute value of

NSLOPEAFD and PSLOPEAFD are the amount of AFD adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14. ** 22f(I)f(I)NSLOPE and PSLOPE are the amounts of the OPT f2(I) limit adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, 95, 131, 146, 206, 214 (NSLOPE % x f(I) margin)2**f(I)2min Page 8 of 10 LA03M05ACTION B REQUIRED ACTION B.2 REQUIRED ACTION B.1 REQUIRED ACTION C.2 ACTION C REQUIRED ACTION B.1/B.2 REQUIRED ACTION C.1/C.2 M05LA03LA03LA03LA03LA03 A01ITS ITS 3.2.1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) (b) Within 48 hours, reduce the OPT positive f2(I) breakpoint limit by: Reduced OPT positive f2(I) breakpoint limit = (f2(I) limit of Table 2.2-1) absolute value of (PSLOPE** % x f(I) margin)2f(I)2min

d. Measuring QMF(X,Y,Z) according to the following schedule: 1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(X,Y,Z) was last determined,*** or 2. At least once per 31 Effective Full Power Days, whichever occurs first. e. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding Z),Y,BQNOM(X, Z)Y,(X,FMQ either of the following actions specified shall be taken.

1. QMF(X,Y,Z) shall be increased over that specified in 4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or 2. QMF(X,Y,Z)shall be evaluated according to 4.2.2.2 at or before the time when the margin is projected to result in one of the actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4. 4.2.2.3 When FQ(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(X,Y,Z) shall be obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the FQ(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

      ** 2f(I)NSLOPE and 2f(I)PSLOPE are the amounts of the OPT f2(I) limit adjustment required to compensate for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14.  *** During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-6a Amendment No. 21, 95, 131, 214 Page 9 of 10 SR Note LA04In accordance with the Surveillance Frequency Control ProgramandM08thereafter M06SR 3.2.1.1 SR 3.2.1.2 SR 3.2.1.3 LA03LA02REQUIRED ACTION C.1 M08REQUIRED ACTION C.1/C.2 LA03M04SR 3.2.1.1 SR 3.2.1.2 / SR 3.2.1.3 NOTE b. SR 3.2.1.2 / SR 3.2.1.3 NOTE a. SR 3.2.1.2/SR 3.2.1.3 NOTE M09INSERT 4 Once within 12 hours after M07A02)Z,Y,X(FCQINSERT 3 Not required to be performed until 12 hours after an equilibrium can be 3.2.1 Insert Page 3/4 2-6a CTS (Page 9 of 10) INSERT 3 Once after each refueling prior to THERMAL POWER exceeding 75% RTP

INSERT 4 Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FCQ was last verified AND At least once per 31 Effective Full Power Days DOC M09 DOC M06 LA04In accordance with the Surveillance Frequency Control Program A01ITS ITS 3.2.1

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March 30, 1992 SEQUOYAH - UNIT 2 3/4 2-7 Amendment No. 131, 146 Page 10 of 10 DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 1 of 9 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 4.2.2.2 evaluates )Z,Y,X(FMQto determine if FQ(X,Y,Z) is within the limits. CTS 4.2.2.3 evaluates FQ(X,Y,Z) for reasons other than meeting the requirements of CTS 4.2.2.2 and requires the overall measured FQ(X,Y,Z) be obtained from a distribution flux map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the FQ(X,Y,Z) limit specified in the COLR. ITS 3.2.1 ACTION A and SR 3.2.1.1 use )Z,Y,X(FCQto represent the overall measured FQ(X,Y,Z) adjusted for measurement uncertainty and manufacturing tolerances. This changes the CTS by adding a new term,)Z,Y,X(FCQ which reflects the requirements in CTS 4.2.2.3 for evaluating the steady state limit of FQ(X,Y,Z) specified in the COLR. BAW-10163PA, "Core Operating Limit Methodology for Westinghouse-Designed PWRs" June 1989, requires that FQ(X,Y,Z) is compared against three limits: (1) steady state limit, (FQRTP / P)

K(Z), (2) limiting condition LOCA limit, BQDES(X,Y,Z), and (3) Limiting condition centerline fuel melt limit, BCDES(X,Y,Z). BAW-10163PA further states that the overall measured FQ(X,Y,Z) must be adjusted for uncertainty prior to comparison to the steady state limit.

The CTS 3.2.2 Surveillance Requirements address both the steady state and the limiting conditions. CTS 4.2.2.2, in part evaluates)Z,Y,X(FMQ for both BQDES(X,Y,Z) and BCDES(X,Y,Z) to ensure the FQ(X,Y,Z) limit is met at limiting conditions. Thus if BQDES(X,Y,Z) and BCDES(X,Y,Z) are met, the steady state limit is met These verifications are reflected in ITS SR 3.2.1.2 and SR 3.2.1.3. CTS 4.2.2.3 addresses evaluation of the steady state limit directly using the overall measured FQ(X,Y,Z) adjusted by the two penalty factors,)Z,Y,X(FCQ. ITS 3.2.1 uses )Z,Y,X(FCQthroughout the Specification to refer to the steady state limit. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 3.2.2 ACTION c states that with FQ(X,Y,Z) exceeding its limit "Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 2 of 9 POWER above the reduced limit required by Action a. and b., above; THERMAL POWER may then be increased provided FQ(X,Y,Z) is demonstrated through incore flux mapping to be within its limits." However, under CTS 3.0.2, the FQ(X,Y,Z) measurement does not have to be completed, if compliance with the LCO is restored. ITS 3.2.1 ACTION A contains a Note which states, "Required Action A.5 must be completed whenever this Condition is entered." ITS Required Action A.5 requires performance of SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 prior to increasing THERMAL POWER above the limit of Required Action A.1. This changes the CTS by requiring )Z,Y,X(FCQverification to be made even if )Z,Y,X(FCQis restored to within its limit. The purpose of CTS 3.2.2 ACTION c is to ensure that when FQ(X,Y,Z) has exceeded the limit, compensatory measures are commenced to restore core power distribution to within the limits prior to increasing THERMAL POWER. This change is acceptable, because it establishes appropriate compensatory measurements for violation of the FQ(X,Y,Z) limit. As power is reduced under ITS Required Action A.1, the margin to the FQ(X,Y,Z) limit increases. Therefore, compliance with the LCO could be restored during the power reduction. Verifying that the limit is met as power is increased ensures that the limit continues to be met and does not remain unmeasured for up to 31 EFPD. This change is designated as a more restrictive change, because it imposes requirements in addition to those in the CTS. M02 CTS 3.2.2 ACTION states in part that when FQ(X,Y,Z) has exceeded the limit, to (1) Reduce THERMAL POWER at least 1% for each 1% FQ(X,Y,Z) exceeds the limit within 15 minutes, (2) Administratively reduce the allowable power at each point along the AFD limit lines within 2 hours, (3) Reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours, (4) Reduce the Overpower Delta T Trip Setpoints (value of K4) at least 1% (in T span) for each 1% that FQ(X,Y,Z) exceeds the limit specified in the COLR within the next 48 hours, (5) Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Action a. and b., above; THERMAL POWER may then be increased provided FQ(X,Y,Z) is demonstrated through incore mapping to be within its limits. ITS 3.2.1 has similar Required Actions and Completion Times with the added requirement to ensure the times are met after each )Z,Y,X(FCQdetermination. This changes the CTS by requiring the Required Actions to be re-performed within a specific Completion Time after each flux map determination. The purpose of the CTS 3.2.2 ACTIONs is to ensure that when FQ(X,Y,Z) has exceeded the limit, compensatory measures are commenced to restore core power distribution to within the limits assumed in the safety analysis. This change is acceptable because it ensures that the Required Actions for )Z,Y,X(FCQ not within limits will be re-performed after each )Z,Y,X(FCQdetermination within the prescribed Completion Time. When )Z,Y,X(FCQ is not met, the margin to the limit prescribes the amount of power reduction and setpoint reduction to be performed. Therefore, each time flux mapping is performed, the determination of margin to the limit will determine if additional power reduction or additional DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 3 of 9 setpoint reduction is required. This change is designated as more restrictive, because it applies new Completion Time requirements which do not exist in the CTS. M03 CTS 3.2.2 does not contain an Action to follow, if the provided Actions cannot be met. Therefore, CTS 3.0.3 would be entered, which would allow 1 hour to initiate a shutdown and to be in HOT STANDBY within 7 hours. ITS 3.2.1 ACTION D, states that the plant must be in MODE 2 within 6 hours, if any Required Action and associated Completion Time is not met. This changes the CTS by eliminating the one hour to initiate a shutdown and, consequently, allowing one hour less for the unit to be in MODE 2. This change is acceptable because it provides an appropriate compensatory measure for the described conditions. If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. The LCO is applicable in MODE 1. Requiring a shutdown to MODE 2 is appropriate in this condition. The one hour allowed by CTS 3.0.3 to prepare for a shutdown is not needed, because the operators have had time to prepare for the shutdown while attempting to follow the Required Actions and associated Completion Times. This change is designated as more restrictive because it allows less time to shut down than does the CTS. M04 CTS 4.2.2.1 states that the provisions of Specification 4.0.4 are not applicable, and thereby provides an allowance for entering the next higher MODE of Applicability when the Surveillance is not met. CTS 4.2.2.2.d.1 Note *** states that during power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained. ITS 3.2.1 has a similar note for the beginning of each cycle, however, there is no specific allowance for changing MODES at any other time with ITS LCO 3.2.1 not met. ITS LCO 3.0.4 requires, in part, that when an LCO is not met, entry into a MODE or other specified condition in the applicability shall only be made: If part a. or part b. or part c. is met. Part c allows, when an allowance is stated in the individual value, parameter or other specification. ITS 3.2.1 Surveillance Requirements Note will provide an allowance whereby, Surveillance performance is not required until 12 hours after an equilibrium power level has been achieved, at which a power distribution map can be obtained. This changes CTS by allowing entry into the MODE of Applicability by only deferring the performance of the Surveillance Requirements instead of deferring compliance with the LCO. The purpose of CTS 4.2.2.1 is to provide an allowance for entering the next higher MODE of applicability when any Surveillance is not met. This change is acceptable because ITS provides an allowance to enter the MODE of Applicability at any time LCO 3.2.1 is not met solely based on Surveillance performance. SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 require using the incore detector system to provide the necessary data to create a power distribution map. To provide the necessary data, MODE 1 needs to be entered, power escalated, stabilized and equilibrium conditions established at some higher power level (~40%-50%). The surveillances cannot be performed in MODE 2. This change is designated as more restrictive because the CTS 4.0.4 MODE DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 4 of 9 change allowance with the LCO not met is now limited to the performance of the SRs and does not include the allowance to change MODES for non-compliance with the acceptance criteria. M05 CTS 3.2.2 provides two acceptable alternatives for the AFD min margin and f2(I) min margin not met. CTS 4.2.2.2.c.3 states, "If the AFD min margin in 4.2.2.2.c.2 above is < 0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed." CTS 4.2.2.2.c.4 states, "If the f2(I) min margin in 4.2.2.2.c.2 above is < 0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed." CTS 4.2.2.2.c.3.a and CTS 4.2.2.2.c.3.b have been replaced by ITS 3.2.1 Required Actions B.1 and B.2. Similarly, CTS 4.2.2.2.c.4.a and 4.2.2.2.c.4.b have been replaced with ITS 3.2.1 Required Actions C.1 and C.2. However, in both cases the option for, "the action statements for 3.2.2 shall be followed" has not been retained. This changes the CTS by removing the option to follow the action statement of CTS 3.2.2 for either min margin (AFD or f2(I)) not met. The purpose of CTS 4.2.2.2.c.3 and CTS 4.2.2.2.c.4 is to provide acceptable alternatives for the required compensatory actions when either AFD min margin or f2(I) min margin is not met. The CTS surveillance requirements for either AFD min margin or f2(I) min margin not met require either the administrative reduction in their respective setpoints or the option of entering the actions of LCO 3.2.2. The CTS Actions for 3.2.2, FQ(X,Y,Z) exceeding the limits, require in part the reduction of THERMAL POWER, reduction of AFD limit lines, and reduction f2(I) breakpoint limits. ITS 3.2.1 has removed this option, but retains the requirement for administrative reduction in AFD limits, ITS CONDITION B, or f2(I) breakpoint limits, ITS CONDITION C. If the ITS Required Actions to administratively reduce the respective setpoints is not performed within the allowed Completion Time, Condition D will be entered requiring the Unit to be placed in MODE 2. This change is designated as more restrictive because an acceptable alternative Required Action available in CTS is being removed.

M06 CTS 4.2.2.2.d requires FQ(X,Y,Z) to be determined to be within its limit upon achieving equilibrium conditions after exceeding by 10 percent or more of RTP, the THERMAL POWER at which FQ(X,Y,Z) was last determined, or at least once per 31 EFPD, whichever occurs first. ITS SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3 collectively verify that FQ(X,Y,Z) is within its limit after each refueling prior to THERMAL POWER exceeding 75% RTP, once within 12 hours after achieving equilibrium conditions after exceeding, by greater than or equal to 10% RTP, the THERMAL POWER at which),,(ZYXFCQand)Z,Y,X(FMQwas last verified, and in accordance with the Surveillance Frequency Control Program. This changes the CTS by adding a new Frequency (Once after each refueling prior to THERMAL POWER exceeding 75% RTP). The replacement of the words "whichever occurs first" with the word "thereafter" to the Frequency is discussed in DOC M08. Moving the "31 EFPD thereafter" Frequency to the Surveillance Frequency Control Program is discussed in DOC LA04. The addition of "once within 12 hours" is discussed in DOC M07. DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 5 of 9 The purpose of SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 is to verify that FQ(X,Y,Z) is within the limits assumed in the safety analysis. This change is acceptable, because adopting the new Frequency of confirming ),,(ZYXFCQand)Z,Y,X(FMQare within the limits prior to exceeding 75% RTP following each core reload, will ensure that some determination of ),,(ZYXFCQand)Z,Y,X(FMQ is made at a lower power level at which adequate margin is available, before going to 100% RTP. This change is designated as more restrictive, because it applies new requirements which do not exist in the CTS. M07 CTS 4.2.2.2.d requires FQ(X,Y,Z) to be determined to be within its limit upon achieving equilibrium conditions after exceeding by 10 percent or more of RTP, the THERMAL POWER at which FQ(X,Y,Z) was last determined, or at least once per 31 EFPD, whichever occurs first. ITS SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3 collectively verify that FQ(X,Y,Z) is within its limit after each refueling prior to THERMAL POWER exceeding 75% RTP, once within 12 hours after achieving equilibrium conditions after exceeding, by greater than or equal to 10% RTP, the THERMAL POWER at which FQ(X,Y,Z) was last verified, and in accordance with the Surveillance Frequency Control Program. This changes the CTS by modifying the existing Frequency (Upon achieving equilibrium conditions-) by adding a specific time element (Once within 12 hours after achieving equilibrium conditions) which limits the time duration allowed for completing a single performance after a 10% RTP change . The replacement of the words "whichever occurs first" with the word "thereafter" to the Frequency is discussed in DOC M08. The relocation of the "31 EFPD thereafter" Frequency to the Surveillance Frequency Control Program is discussed in DOC LA04. The addition of new Frequency (Once after each refueling prior to THERMAL POWER exceeding 75% RTP) is discussed in DOC M06. The purpose of SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 is to verify that FQ(X,Y,Z) is within the limits assumed in the safety analysis. This change is acceptable, because modifying the existing frequency by adding a specific time element completing a single performance after a 10% RTP change is made ensures adequate margin is available, before going to a higher power level. This change is designated as more restrictive, because it applies new requirements which do not exist in the CTS. M08 CTS 4.2.2.2.d.1 Surveillance states "required to be performed upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(X,Y,Z) was last determined, or at least once per 31 EFPD, whichever occurs first." ITS SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 are similar, but the logical connector between the Frequencies is an "AND" not an "or". Additionally, the ITS 31 EFPD Frequency is qualified with "thereafter". This changes the CTS by (1) removing the phrase, "whichever occurs first" and replacing it with "thereafter" and (2) changing the CTS logical connector from "or" to "AND". The purpose of CTS 4.2.2.2 is to establish both when and how often )Z,Y,X(FMQis measured. The intent of the CTS Frequency logical connector "or" does not provide an exclusion to perform either the situational performance or the DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 6 of 9 repetitive performance of the test, because both are continuously applicable when )Z,Y,X(FMQis measured. Additionally, the CTS Frequency describes "when" the first performance is required (i.e. whichever occurs first) based on plant conditions. This change is acceptable because the ITS use of "AND" will ensure both the situational and periodic performances are continuously applicable. This change is designated more restrictive because the Surveillance Requirements will be required to be performed more frequently than is required in CTS. M09 CTS 4.2.2.3 states that when FQ(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(X,Y,Z) shall be obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the FQ(X,Y,Z) limit specified in the COLR according to Specification 3.2.2. Proposed ITS SR 3.2.1.1, verifies )Z,Y,X(FCQis within the steady state limits, (1) Once after each refueling prior to THERMAL POWER exceeding 75% RTP, and (2) Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FCQwas last verified, and (3) In accordance with the Surveillance Frequency Control Program. This changes the CTS from a 4.2.2.3 measurement of FQ(X,Y,Z) to be within limits on a situational Frequency basis to the ITS Frequency of (1) Once after each refueling prior to THERMAL POWER exceeding 75% RTP, and (2) Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FCQwas last verified, and (3) In accordance with the Surveillance Frequency Control Program. (The relocation of "31 EFPD, thereafter to the Surveillance Frequency Control program" is discussed in DOC LA04 (Refer to DOC A02 for the discussion of the addition of a new term describing the steady state limit,)Z,Y,X(FCQ). The purpose of CTS 4.2.2.3 is to evaluate FQ(X,Y,Z) during those situational conditions where core power distribution limits may exceed limits assumed in the safety analysis. BAW-10163PA "Core Operating Limit Methodology for Westinghouse-Designed PWRs" June 1989 requires the measured FQ(X,Y,Z) to be compared against the steady state limit (ITS SR 3.2.1.1) and the two transient limits BQDES(X,Y,Z)(ITS SR 3.2.1.2) and BCDES(X,Y,Z)(ITS SR 3.2.1.3). ITS SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 will be performed at the same Frequencies. This change is designated as more restrictive because the situational testing Frequency of CTS 4.2.2.3 is being replaced with two new situational Frequencies and a periodic performance, once every 31 EFPD.

RELOCATED SPECIFICATIONS None DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 7 of 9 REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.2.2 ACTION b requires within 48 hours when FQ(X,Y,Z) is not within limits to reduce the Overpower Delta T Trip setpoints (value of K4) at least 1% (in T span) for each 1% that FQ(X,Y,Z) exceeds the limit provided in the COLR. ITS LCO 3.2.1 Required Action A.3 requires within 48 hours of discovery that )Z,Y,X(FCQis not within limits, to reduce Overpower T trip setpoints at least 1% for each 1% that )Z,Y,X(FCQ exceeds the limit. This changes the CTS by moving the specific information regarding the terms, "value of K4" and "in T span," to the COLR. The removal of these details for performing actions from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements to reduce Overpower T trip setpoints at least 1% for each 1% that )Z,Y,X(FCQexceeds the limit. Also, this change is acceptable because the removed information will be adequately controlled in the COLR requirements provided in ITS 5.6.5, "Core Operating Limits Report." ITS 5.6.5 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such as transient analysis limits and accident analysis limits) of the safety analyses are met. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.2.2 ACTION c requires FQ(X,Y,Z) to be determined to be within its limit through incore mapping. CTS 4.2.2.3 requires FQ(X,Y,Z) to be determined to be within its limit by obtaining a power distribution map and applying manufacturing tolerances and measurement uncertainty factors before comparing the results to the FQ(X,Y,Z) limit specified in the COLR. ITS 3.2.1 Required Action A.5 and ITS SR 3.2.1.1 require verification that )Z,Y,X(FCQ is within its limit. This changes the CTS by moving the manner in which the FQ(X,Y,Z) determination is performed to the Bases. The removal of these details for performing actions from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to determine FQ(X,Y,Z) is within its limit. Also, this change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 8 of 9 LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.2.2.2, 4.2.2.2.a, 4.2.2.2.b, 4.2.2.2.*, 4.2.2.2.**, 4.2.2.2.c.1, 4.2.2.2.c.2, 4.2.2.2.c.3.a, 4.2.2.2.c.3.b, 4.2.2.2.c.4.a, 4.2.2.2.c.4.b, 4.2.2.2.d, and 4.2.2.3 provide details for evaluating FMQ(X,Y,Z) to determine if FQ(X,Y,Z) is within limits. ITS SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 collectively verify that FQ(X,Y,Z) (as discussed in DOC A4) is within limits specified in the COLR. This changes the CTS by moving the details for evaluating FMQ(X,Y,Z) to determine if FQ(X,Y,Z) is within limits to the ITS Bases. The removal of these details from the Technical Specifications and their relocation into the ITS Bases is acceptable, because the procedural steps and further details for making a determination that FQ(X,Y,Z) is within its limits is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS Surveillance Requirement to verify FQ(X,Y,Z) is within its limits will more closely align with the LCO requirement for FQ(X,Y,Z) to be within the limits specified in the COLR. Also, this change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA04 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.2.2.2 requires, in part, a determination that FQ(X,Y,Z) is within its limits at least once per 31 EFPD. ITS SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.1.3 require a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 9 of 9 LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.2.2 ACTION a.2 states, in part, that when FQ(X,Y,Z) exceeds its limit, reduce the Power Range Neutron Flux - High Trip setpoints within the next 4 hours. ITS 3.2.1 Required Actions A.4 states with )Z,Y,X(FCQ not within limit, reduce the Power Range Neutron Flux - High Trip setpoints by 1% for each 1% )Z,Y,X(FCQ exceeds the limit. The ITS 3.2.1 Required Action A.4 Completion Time is "within 72 hours after each )Z,Y,X(FCQdetermination." This changes the CTS by increasing the time allowed to reduce the trip setpoints. The purpose of CTS 3.2.2 ACTION a.2 is to lower the Power Range Neutron Flux - High Trip setpoints, which ensures continued operation is at an acceptably low power level with an adequate DNBR margin and avoids violating the )Z,Y,X(FCQ limit. This change is acceptable, because the Completion Time is consistent with safe operation and recognizes that the safety analysis assumptions are satisfied once power is reduced, and considers the low probability of a DBA occurring during the allowed Completion Time. The revised Completion Time allows the Power Range Neutron Flux - High Trip setpoints to be reduced in a controlled manner without challenging operators, technicians, or plant systems. Following a significant power reduction, a time period of 24 hours is allowed to reestablish steady state xenon concentration and power distribution and to take and analyze a flux map. If it is determined that )Z,Y,X(FCQis still not within its limit, reducing the Power Range Neutron Flux - High Trip Setpoints can be accomplished within a few hours. Furthermore, setpoint changes should only be required for extended operation in this condition, because of the risk of a plant trip during the adjustment. This change is designated as less restrictive, because additional time is allowed to lower the Power Range Neutron Flux - High Trip setpoints than was allowed in the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-1 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 1 Amendment xxx 1223.2 POWER DISTRIBUTION LIMITS 3.2.1B Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology) LCO 3.2.1B FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, shall be within the limits specified in the COLR. APPLICABILITY: MODE 1. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ Required Action A.4 shall be completed whenever this Condition is entered. ---------------------------------

)Z(FCQ not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1% )Z(FCQ exceeds limit. AND A.2 Reduce Power Range Neutron Flux - High trip setpoints 1% for each 1% )Z(FCQ exceeds limit. AND A.3 Reduce Overpower T trip setpoints 1% for each 1% )Z(FCQ exceeds limit. AND A.4 Perform SR 3.2.1.1 and SR 3.2.1.2. 15 minutes after each )Z(FCQ determination

72 hours after each )Z(FCQ determination

Prior to increasing THERMAL POWER above the limit of Required Action A.1 X,Y,Z 22X,Y,Z 3.2.2 Applicability ACTION a DOC L01 ACTION a.2 DOC M02 ACTION b DOC M02 ACTION c DOC M02 ),,(ZYXFCQ 111411DOC M01 4548 INSERT 1 the steady state 5 ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ, SR 3.2.1.23DOC M02 33333 3.2.1 Insert Page 3.2.1-1 CTS INSERT 1 CONDITION REQUIRED ACTION COMPLETION TIME AND A.2 Reduce, by administrative means, AFD limits 1% for each 1% )Z,Y,X(FCQ exceeds limit. 2 hours after each )Z,Y,X(FCQ determination 3ACTION a.1 DOC M02 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-2 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 1 Amendment xxx 122ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE------------ Required Action B.4 shall be completed whenever this Condition is entered. --------------------------------- )Z(FWQ not within limits. B.1 Reduce AFD limits 1% for each 1% )Z(FWQ exceeds limit. AND B.2 Reduce Power Range Neutron Flux - High trip setpoints 1% for each 1% that the maximum allowable power of the AFD limits is reduced. AND B.3 Reduce Overpower T trip setpoints 1% for each 1% that the maximum allowable power of the AFD limits is reduced. AND B.4 Perform SR 3.2.1.1 and SR 3.2.1.2. 4 hours 72 hours

72 hours

Prior to increasing THERMAL POWER above the maximum allowable power of the AFD limits C. Required Action and associated Completion Time not met. C.1 Be in MODE 2. 6 hours 4D D DOC M03 INSERT 2 3.2.1 Insert Page 3.2.1-2 CTS INSERT 2 CONDITION REQUIRED ACTION COMPLETION TIME B. AFD min margin < 0 B.1 Reduce, by administrative means, positive AFD limit lines for each power level by PSLOPEAFD for each 1% FQ(X,Y,Z) exceeds limit. AND B.2 Reduce, by administrative means, negative AFD limit lines for each power level by NSLOPEAFD for each 1% FQ(X,Y,Z) exceeds limit. 2 hours

2 hours C. f2(I) min margin < 0 C.1 Reduce Overpower T positive f2(I) breakpoint limit by PSLOPEf2(I) for each 1% FQ(X,Y,Z) exceeds limit. AND C.2 Reduce Overpower T negative f2(I) breakpoint limit by NSLOPEf2(I) for each 1% FQ(X,Y,Z) exceeds limit. 48 hours

48 hours 44.2.2.2.c.3.b Note*4.2.2.2.c.3.a Note

4.2.2.2.c.4.b Note ** 4.2.2.2.c.4.a Note ** 4.2.2.2.c.3 4.2.2.2.c.3 4.2.2.2.c.4 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-3 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 1 Amendment xxx 122SURVEILLANCE REQUIREMENTS ------------------------------------------------------------NOTE----------------------------------------------------------- During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify )Z(FCQ is within limit.

Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within [12] hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z(FCQ was last verified

AND [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4.2.2.2 Note *** 4.2.2.2 4.2.2.3 DOC M06 DOC M09 DOC A02 117884.2.2.2.d.2 4.2.2.2.d.1 DOC M07 the steady state),,(ZYXFCQ),,(ZYXFCQ4.2.2.1 DOC M04 5can be5INSERT 3 6Not required to be performed until 12 hours after 3.2.1 Insert Page 3.2.1-3 CTS INSERT 3 -----------------------------------------NOTE------------------------------------ Not required to be performed if SR 3.2.1.2 and SR 3.2.1.3 are met. --------------------------------------------------------------------------------------

64.2.2.2 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-4 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 1 Amendment xxx 122SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.2.1.2 -------------------------------NOTE------------------------------ If measurements indicate that the maximum over z [ )Z(FCQ / K(Z) ] has increased since the previous evaluation of )Z(FCQ:

a. Increase )Z(FWQ by the greater of a factor of [1.02] or by an appropriate factor specified in the COLR and reverify )Z(FWQ is within limits or b. Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [)Z(FCQ / K(Z) ] has not increased. ---------------------------------------------------------------------

Verify )Z(FWQ is within limit.

Once after each refueling prior to THERMAL POWER exceed-ing 75% RTP AND Once within [12] hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z(FWQ was last verified AND 4 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-5 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 1 Amendment xxx 122SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4INSERT 4 4 3.2.1 Insert Page 3.2.1-5a CTS INSERT 4 SURVEILLANCE FREQUENCY SR 3.2.1.2 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: )Z,Y,X(FMQ > BQNOM(X,Y,Z) a.Increase )Z,Y,X(FMQ by the appropriate factor specified in the COLR and reverifyAFD min margin > 0; orb.Repeat SR 3.2.1.2 prior to the time at which theprojected AFD min margin will be < 0.--------------------------------------------------------------------- Verify AFD min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FMQ was last verified AND 44.2.2.2.e 4.2.2.2.e.1 4.2.2.2.e.2 4.2.2.2.c.1 DOC M06 4.2.2.2.d.1 DOC M07 3.2.1 Insert Page 3.2.1-5b CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY [ 31 EFPD thereafter] OR In accordance with the Surveillance Frequency Control Program ] 4.2.2.2.d.2 4 3.2.1 Insert Page 3.2.1-5c CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY SR 3.2.1.3 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: )Z,Y,X(FMQ > BQNOM(X,Y,Z) a.Increase )Z,Y,X(FMQ by the appropriate factor specified in the COLR and reverifyf2(I) min margin > 0; orb.Repeat SR 3.2.1.3 prior to the time at which theprojected f2(I) min margin will be < 0.--------------------------------------------------------------------- Verify f2(I) min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FMQ was last verified AND 4 4.2.2.2.e.1 4.2.2.2.e.2 4.2.2.2.e 4.2.2.2.c.1 DOC M06 4.2.2.2.d.1 DOC M07 3.2.1 Insert Page 3.2.1-5d CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4.2.2.2.d.2 4 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS 3.2.1B-1 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 2 Amendment xxx 1223.2 POWER DISTRIBUTION LIMITS 3.2.1B Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology) LCO 3.2.1B FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, shall be within the limits specified in the COLR. APPLICABILITY: MODE 1. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ Required Action A.4 shall be completed whenever this Condition is entered. ---------------------------------

)Z(FCQ not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1% )Z(FCQ exceeds limit. AND A.2 Reduce Power Range Neutron Flux - High trip setpoints 1% for each 1% )Z(FCQ exceeds limit. AND A.3 Reduce Overpower T trip setpoints 1% for each 1% )Z(FCQ exceeds limit. AND A.4 Perform SR 3.2.1.1 and SR 3.2.1.2. 15 minutes after each )Z(FCQ determination

72 hours after each )Z(FCQ determination

Prior to increasing THERMAL POWER above the limit of Required Action A.1 X,Y,Z 22X,Y,Z 3.2.2 Applicability ACTION a DOC L01 ACTION a.2 DOC M02 ACTION b DOC M02 ACTION c DOC M02 ),,(ZYXFCQ 111411DOC M01 4548 INSERT 1 the steady state 5 ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ),,(ZYXFCQ, SR 3.2.1.23DOC M02 33333 3.2.1 Insert Page 3.2.1-1 CTS INSERT 1 CONDITION REQUIRED ACTION COMPLETION TIME AND A.2 Reduce, by administrative means, AFD limits 1% for each 1% )Z,Y,X(FCQ exceeds limit. 2 hours after each )Z,Y,X(FCQ determination 3ACTION a.1 DOC M02 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS3.2.1B-2 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 2 Amendment xxx 122ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE------------ Required Action B.4 shall be completed whenever this Condition is entered. --------------------------------- )Z(FWQ not within limits. B.1 Reduce AFD limits 1% for each 1% )Z(FWQ exceeds limit. AND B.2 Reduce Power Range Neutron Flux - High trip setpoints 1% for each 1% that the maximum allowable power of the AFD limits is reduced. AND B.3 Reduce Overpower T trip setpoints 1% for each 1% that the maximum allowable power of the AFD limits is reduced. AND B.4 Perform SR 3.2.1.1 and SR 3.2.1.2. 4 hours 72 hours 72 hours Prior to increasing THERMAL POWER above the maximum allowable power of the AFD limits C. Required Action and associated Completion Time not met. C.1 Be in MODE 2. 6 hours 4D D DOC M03 INSERT 2 3.2.1 Insert Page 3.2.1-2 CTS INSERT 2 CONDITION REQUIRED ACTION COMPLETION TIME B. AFD min margin < 0 B.1 Reduce, by administrative means, positive AFD limit lines for each power level by PSLOPEAFD for each 1% FQ(X,Y,Z) exceeds limit. AND B.2 Reduce, by administrative means, negative AFD limit lines for each power level by NSLOPEAFD for each 1% FQ(X,Y,Z) exceeds limit. 2 hours

2 hours C. f2(I) min margin < 0 C.1 Reduce Overpower T positive f2(I) breakpoint limit by PSLOPEf2(I) for each 1% FQ(X,Y,Z) exceeds limit. AND C.2 Reduce Overpower T negative f2(I) breakpoint limit by NSLOPEf2(I) for each 1% FQ(X,Y,Z) exceeds limit. 48 hours

48 hours 44.2.2.2.c.3.b Note*4.2.2.2.c.3.a Note

4.2.2.2.c.4.b Note ** 4.2.2.2.c.4.a Note ** 4.2.2.2.c.3 4.2.2.2.c.3 4.2.2.2.c.4 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS3.2.1B-3 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 2 Amendment xxx 122SURVEILLANCE REQUIREMENTS ------------------------------------------------------------NOTE----------------------------------------------------------- During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify )Z(FCQ is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within [12] hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z(FCQ was last verified AND [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4.2.2.2 Note *** 4.2.2.2 4.2.2.3 DOC M06 DOC M09 DOC A02 117884.2.2.2.d.2 4.2.2.2.d.1 DOC M07 the steady state),,(ZYXFCQ),,(ZYXFCQ4.2.2.1 DOC M04 5can be5INSERT 3 6Not required to be performed until 12 hours after 3.2.1 Insert Page 3.2.1-3 CTS INSERT 3 -----------------------------------------NOTE------------------------------------ Not required to be performed if SR 3.2.1.2 and SR 3.2.1.3 are met. -------------------------------------------------------------------------------------- 64.2.2.2 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS3.2.1B-4 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 2 Amendment xxx 122SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.2.1.2 -------------------------------NOTE------------------------------ If measurements indicate that the maximum over z [ )Z(FCQ / K(Z) ] has increased since the previous evaluation of )Z(FCQ: a.Increase )Z(FWQ by the greater of a factor of[1.02] or by an appropriate factor specified inthe COLR and reverify )Z(FWQ is within limits or b.Repeat SR 3.2.1.2 once per 7 EFPD until eithera.above is met or two successive flux mapsindicate that the maximum over z [)Z(FCQ / K(Z) ] has not increased. --------------------------------------------------------------------- Verify )Z(FWQ is within limit. Once after each refueling prior to THERMAL POWER exceed-ing 75% RTP AND Once within [12] hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z(FWQ was last verified AND 4 FQ(Z) (RAOC-W(Z) Methodology) 3.2.1B WOG STS3.2.1B-5 Rev. 4.0, CTS X,Y,Z 1SEQUOYAH UNIT 2 Amendment xxx 122SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 4INSERT 4 4 3.2.1 Insert Page 3.2.1-5a CTS INSERT 4 SURVEILLANCE FREQUENCY SR 3.2.1.2 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: )Z,Y,X(FMQ > BQNOM(X,Y,Z) a.Increase )Z,Y,X(FMQ by the appropriate factor specified in the COLR and reverifyAFD min margin > 0; orb.Repeat SR 3.2.1.2 prior to the time at which theprojected AFD min margin will be < 0.--------------------------------------------------------------------- Verify AFD min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FMQ was last verified AND 44.2.2.2.e 4.2.2.2.e.1 4.2.2.2.e.2 4.2.2.2.c.1 DOC M06 4.2.2.2.d.1 DOC M07 3.2.1 Insert Page 3.2.1-5b CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY

[ 31 EFPD thereafter] OR In accordance with the Surveillance Frequency Control Program ] 4.2.2.2.d.2 4 3.2.1 Insert Page 3.2.1-5c CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY SR 3.2.1.3 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: )Z,Y,X(FMQ > BQNOM(X,Y,Z) a.Increase )Z,Y,X(FMQ by the appropriate factor specified in the COLR and reverifyf2(I) min margin > 0; orb.Repeat SR 3.2.1.3 prior to the time at which theprojected f2(I) min margin will be < 0.--------------------------------------------------------------------- Verify f2(I) min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which )Z,Y,X(FMQ was last verified AND 4 4.2.2.2.e.1 4.2.2.2.e.2 4.2.2.2.e 4.2.2.2.c.1 DOC M06 4.2.2.2.d.1 DOC M07 3.2.1 Insert Page 3.2.1-5d CTS INSERT 4 (continued) SURVEILLANCE FREQUENCY [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ]

4.2.2.2.d.2 4 JUSTIFICATION FOR DEVIATIONS ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 1 of 2 1.Changes are made (additions, deletions, and/or changes) to the ISTS whichreflect the plant specific nomenclature, number, reference, system description,analysis, or licensing basis description.2.The RAOC-W(Z) methodology and the Specification designator "B" are deletedbecause they are unnecessary. (Only one Heat Flux Hot Channel Factor Specification is used in the SQN ITS). This information is provided in NUREG-1431, Rev. 4 to assist in identifying the appropriate Specification to be used as a model for the plant specific ITS conversion, but serves no purpose in a plant specific implementation. In addition, the CAOC-FXY and CAOC-W(Z)methodology Specifications (ISTS 3.2.1A and 3.2.1C) are not used and are not shown.3.ISTS ACTIONS do not contain a requirement to reduce the AFD limits whenACTION A is entered for )Z,Y,X(FCQnot met. CTS 3.2.2 ACTION a.1 requires a reduction of the allowable power at each point along the AFD limit lines to be reduced within 2 hours. This requirement and Completion Time are being added as Required Action A.2.4.ISTS SR 3.2.1.2 and ISTS ACTION B have been deleted. CTS does not includerequirements to verify )Z(FWQis within limits, or actions to take if )Z(FWQis not within limits. However, CTS does require the verification that both AFD min margin is > 0 and f2(I) min margin is > 0. Additionally, the CTS specifies theactions to take if the above verifications are not met. These verifications and actions are added to ITS 3.2.1 as SR 3.2.1.2 and SR 3.2.1.3 with the associated ACTIONS B and C.5.ISTS 3.2.1 Surveillance Requirements Note allows, during power escalation atthe beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map is obtained. CTS 3.2.2 *** Note has a similar allowance. However, in both CTS and ISTS the allowance is for the first power escalation at the beginning of a new core cycle. Additionally, the CTS has SR 4.2.2.1 which provides, The provisions of Specification 4.0.4 are not applicable. This allowance enables SQN to enterthe MODE of Applicability with the Surveillance not being met. ISTS does nothave a similar allowance in LCO 3.2.1. Therefore, SQN is retaining the allowance to change the MODE of Applicability with the surveillance not being met by modifying the existing Surveillance Note.6.ISTS 3.2.1.1 has been modified by a Note providing an allowance to not performSR 3.2.1.1 if the Surveillance has been determined to be met based on the performance results of both SR 3.2.1.2 and SR 3.2.1.3. If both the AFD min margin and the f2(I) min margin are positive, then the steady state limit is metbecause these margins represent bounding limiting conditions.7.The ISTS contains bracketed information and/or values that are generic to allWestinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. JUSTIFICATION FOR DEVIATIONS ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 2 of 2 8. ISTS SR 3.2.1.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-1 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12B 3.2 POWER DISTRIBUTION LIMITS

B 3.2.1B Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology) BASES BACKGROUND The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core. FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core. During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO(QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis. FQ(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution. FQ(Z) is measured periodically using the incore detector system. These measurements are generally taken with the core at or near equilibrium conditions.

Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ(Z) which are present during nonequilibrium situations such as load following or power ascension.

To account for these possible variations, the equilibrium value of FQ(Z) is adjusted as )Z(FWQby an elevation dependent factor that accounts for the calculated worst case transient conditions. Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. X,Y,Z 21X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z 1111and by assembly location, X, Y INSERT 1 8 3.2.1 Insert Page B 3.2.1-1 INSERT 1 "the FQ(X,Y,Z) limits, BQDES(X,Y,Z) and BCDES(X,Y,Z), have been adjusted by pre-calculated factors (MQ(X,Y,Z) and MC(X,Y,Z) respectively) to account for the calculated worst case transient conditions."1 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-2 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES

APPLICABLE This LCO precludes core power distributions that violate the following SAFETY fuel design criteria: ANALYSES a. During a large break loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200°F (Ref. 1), b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition,

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2), and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting. FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents

FQ(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships: FQ(Z) (CFQ / P) K(Z) for P > 0.5 FQ(Z) (CFQ / 0.5) K(Z) for P 0.5 where: CFQ is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) as a function of core height provided in the COLR, and P = THERMAL POWER / RTP 1X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z RTPQFRTPQFFQ (X,Y,Z)X,Y,Z RTPQF 1.8 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-3 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES

LCO (continued) For this facility, the actual values of CFQ and K(Z) are given in the COLR; however, CFQ is normally a number on the order of [2.32], and K(Z) is a function that looks like the one provided in Figure B 3.2.1B-1. For Relaxed Axial Offset Control operation, FQ(Z) is approximated by )Z(FCQ and )Z(FWQ. Thus, both )Z(FCQ and )Z(FWQ must meet the preceding limits on FQ(Z). An )Z(FCQ evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value ()Z(FMQ) of FQ(Z). Then, )Z(FCQ = )Z(FMQ [1.0815] where [1.0815] is a factor that accounts for fuel manufacturing tolerances and flux map measurement uncertainty. )Z(FCQ is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the incore flux map was taken. The expression for )Z(FWQ is: )Z(FWQ= )Z(FCQW(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR. The )Z(FCQ is calculated at equilibrium conditions. The FQ(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA. This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If )Z(FCQ cannot be maintained within the LCO limits, reduction of the core power is required and if )Z(FWQcannot be maintained within the LCO limits, reduction of the AFD limits is required. Note that sufficient reduction of the AFD limits will also result in a reduction of the core power. Violating the LCO limits for FQ(Z) produces unacceptable consequences if a design basis event occurs while FQ(Z) is outside its specified limits. RTPQFRTPQF2.62134INSERT 2 X,Y,Z )Z,Y,X(FCQX,Y,Z X,Y,Z 115SQN 3.2.1 Insert Page B 3.2.1-3a INSERT 2 Measured FQ(X,Y,Z) is compared against three limits:

Steady state limit, (FQRTP / P)
K(Z),
Limiting condition LOCA limit, BQDES(X,Y,Z), and
Limiting condition centerline fuel melt limit, BCDES(X,Y,Z). FQ(X,Y,Z) is approximated by )Z,Y,X(FCQ for the steady state limit on FQ(X,Y,Z). An )Z,Y,X(FCQ evaluation requires using the moveable incore detectors to obtain a power distribution map in MODE 1. From the incore flux map results we obtain the measured value ()Z,Y,X(FMQ) of FQ(X,Y,Z). Then, )Z,Y,X(FCQ = overall measured FQ(X,Y,Z)
1.05
1.03 where, 1.05 is the measurement reliability factor that accounts for flux map measurement uncertainty (Reference 5) and 1.03 is the local engineering heat flux hot channel factor to account for fuel rod manufacturing tolerance (Reference 4).

BQDES(X,Y,Z) and BCDES(X,Y,Z) are cycle dependent design limits to ensure the FQ(X,Y,Z) limit is met during transients. An evaluation of these limits requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the assembly nodal measured value ()Z,Y,X(FMQ) of FQ(X,Y,Z). )Z,Y,X(FMQ is compared directly to the limits BQDES(X,Y,Z) and BCDES(X,Y,Z). This is appropriate since BQDES(X,Y,Z) and BCDES(X,Y,Z) have been adjusted for uncertainties. The expression for BQDES(X,Y,Z) is: BQDES(X,Y,Z) = Pd(X,Y,Z)

MQ(X,Y,Z)
NRF
F1 / MRF where:
BQDES(X,Y,Z) is the cycle dependent maximum allowable design peaking factor for fuel assembly X,Y at axial location Z. BQDES(X,Y,Z) ensures that the LOCA limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties;
Pd(X,Y,Z) is the design power distribution for fuel assembly X,Y at axial location Z, including the operational flexibility factor;
MQ(X,Y,Z) is the minimum available margin ratio for the LOCA limit at assembly X,Y and axial location Z;
NRF is the nuclear reliability factor;
F1 is the spacer grid factor;
MRF is measurement reliability factor. 4 3.2.1 Insert Page B 3.2.1-3b INSERT 2 (continued) The expression for BCDES(X,Y,Z) is: BCDES(X,Y,Z) = Pd(X,Y,Z)
MC(X,Y,Z)
NRF
F1 / MRF where:
BCDES(X,Y,Z) is the cycle dependent maximum allowable design peaking factor for fuel assembly X,Y, at axial location Z. BCDES(X,Y,Z) ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties;
MC(X,Y,Z) is the minimum available margin ratio for the centerline fuel melt limit at assembly X,Y and axial location Z; The reactor core is operating as designed if the measured steady state core power distribution agrees with prediction within statistical variation. This guarantees that the operating limits will preserve the thermal criteria in the applicable safety analyses. The core is operating as designed if the following relationship is satisfied: )Z,Y,X(FMQ BQNOM(X,Y,Z) where:
BQNOM(X,Y,Z) is the nominal design peaking factor for assembly X,Y at axial location Z increased by an allowance for the expected deviation between the measured and predicted design power distribution. The FQ(X,Y,Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA. BQNOM (X,Y,Z), BQDES(X,Y,Z), and BCDES(X,Y,Z) Data bases are provided for the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in the COLR.

4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-4 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES

APPLICABILITY The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses. Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power. ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which )Z(FCQ exceeds its limit, maintains an acceptable absolute power density. )Z(FCQ is )Z(FMQmultiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. )Z(FMQ is the measured value of FQ(Z). The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of )Z(FCQ and would require power reductions within 15 minutes of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable power level. Decreases in )Z(FCQ would allow increasing the maximum allowable power level and increasing power up to this revised limit.

A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1% by which )Z(FCQ exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of )Z(FCQ and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. Decreases in )Z(FCQ would allow increasing the maximum allowable Power Range Neutron Flux - High trip setpoints. X,Y,Z 15)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FMQ 5)Z,Y,X(FMQ FQ (X,Y,Z))Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ4 )Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQINSERT 3 54 Move to next page after A.3 3.2.1 Insert Page B 3.2.1-4 INSERT 3 A.2 Required Action A.2 requires an administrative reduction of the AFD limits by 1% for each 1% by which )Z,Y,X(FCQ exceeds the steady state limit. The allowed Completion Time of 2 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factor limits are not exceeded. The maximum allowable AFD limits initially determined by Required Action A.2 may be affected by subsequent determinations of )Z,Y,X(FCQ and would require further AFD limit reductions within 2 hours of the )Z,Y,X(FCQdetermination, if necessary to comply with the decreased maximum allowable AFD limits. Decreases in),,(ZYXFCQwould allow increasing the maximum allowable AFD limits. 5 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-5 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES ACTIONS (continued) A.3 Reduction in the Overpower T trip setpoints (value of K4) by 1% for each 1% by which )Z(FCQ exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower T trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of )Z(FCQ and would require Overpower T trip setpoint reductions within 72 hours of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable Overpower T trip setpoints. Decreases in )Z(FCQ would allow increasing the maximum allowable Overpower T trip setpoints. A.4 Verification that )Z(FCQ has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions. Condition A is modified by a Note that requires Required Action A.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.

B.1 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, )Z(FWQ, exceeds its specified limits, there exists a potential for )Z(FCQ to become excessively high if a normal operational transient occurs. Reducing the AFD by 1% for each 1% by which )Z(FWQ exceeds its limit within the allowed Completion Time of 4 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factors are not exceeded. 51555154848)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ3 , SR 3.2.1.2 INSERT 4 INSERT 6 INSERT 5 in T span5543steady state and transient 55 , SR 3.2.1.2 , SR 3.2.1.2 35 )Z,Y,X(FCQ)Z,Y,X(FCQ 3.2.1 Insert Page B 3.2.1-5a INSERT 4 Since)Z,Y,X(FCQ exceeds the steady state limit, the limiting condition operational limit (BQDES) and the limiting condition Reactor Protection System limit (BCDES) may also be exceeded. By performing SR 3.2.1.2 and SR 3.2.1.3, appropriate actions with respect to reductions in AFD limits and OPT trip setpoints will be performed, ensuring that core conditions during operational and Condition II transients are maintained within the bounds of the safety analysis. 5 3.2.1 Insert Page B 3.2.1-5b INSERT 5 B.1 and B.2 The FQ(X,Y,Z) margin supporting AFD operational limits (AFD margin) during transient operations is based on the relationship between)Z,Y,X(FMQ and the limiting condition operational limit, BQDES (X,Y,Z), as follows:

%AFD margin = %*),,(),,(1001ZYXBQDESZYXFMQ The AFD min margin = minimum % margin value of all locations examined. If the reactor core is operating as designed, then )Z,Y,X(FMQ is less than BQDES (X,Y,Z) and calculation of %AFD margin is not required. If the AFD margin is less than zero, then )Z,Y,X(FMQ is greater than BQDES (X,Y,Z) and the AFD limits may not be adequate to prevent exceeding the peaking criteria for a LOCA if a normal operational transient occurs.

Required Actions B.1 and B.2 require reducing the AFD limit lines as follows. The AFD limit reduction is from the full power AFD limits. The adjusted AFD limits must be used until a new measurement shows that a smaller adjustment can be made to the AFD limits, or that no adjustment is necessary:

APL = PAFDL - Absolute Value of (PSLOPEAFD * % AFD Margin) ANL = NAFDL + Absolute Value of (NSLOPEAFD * % AFD Margin) where,

APL is the adjusted positive AFD limit.
ANL is the adjusted negative AFD limit.
PAFDL is the positive AFD limit defined in the COLR.
NAFDL is the negative AFD limit defined in the COLR.
PSLOPEAFD is the adjustment to the positive AFD limit required to compensate for each 1% that)Z,Y,X(FMQ exceeds BQDES (X,Y,Z) as defined in the COLR.
NSLOPEAFD is the adjustment to the negative AFD limit required to compensate for each 1% that)Z,Y,X(FMQ exceeds BQDES (X,Y,Z) as defined in the COLR. * % AFD Margin is the most negative margin determined above. Completing Required Actions B.1 and B.2 within the allowed Completion Time of 2 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factor limits are not exceeded. 54 3.2.1 Insert Page B 3.2.1-5c INSERT 6 C.1 and C.2 The FQ(X,Y,Z) margin supporting the Overpower T f2(I) breakpoints (f2(I) margin) during transient operations is based on the relationship between )Z,Y,X(FMQ and the limit, BCDES(X,Y,Z), as follows: % f2(I) margin = %*),,(),,(1001ZYXBCDESZYXFMQ The f2(I) min margin = minimum % margin value of all locations examined. If the reactor core is operating as designed, then )Z,Y,X(FMQ is less than BCDES(X,Y,Z) and calculation of % f2(I) margin is not required. If the f2(I) margin is less than zero, then )Z,Y,X(FMQ is greater than BCDES(X,Y,Z) and there is a potential that the f2(I) limits are insufficient to preclude centerline fuel melt during certain transients.

Required Actions C.1 and C.2 require reducing the f2(I) breakpoint limits as follows. The f2(I) breakpoint limit reduction is always from the full power f2(I) breakpoint limits. The adjusted f2(I) breakpoint limits must be used until a new measurement shows that a smaller adjustment can be made to the f2(I) breakpoint limits, or that no adjustment is necessary. Posf2(I)Adjusted = Posf2(I)Limit - Absolute Value of (PSLOPEf2(I) * % f2(I) Margin) Negf2(I)Adjusted = Negf2(I)Limit + Absolute Value of (NSLOPEf2(I) * % f2(I) Margin) where:

Posf2(I)Adjusted is the adjusted OPT positive f2(I) breakpoint limit.
Negf2(I)Adjusted is the adjusted OPT negative f2(I) breakpoint limit.
Posf2(I)Limit is the OPT positive f2(I) breakpoint limit defined in the COLR.
Negf2(I)Limit is the OPT negative f2(I) breakpoint limit defined in the COLR.
PSLOPEf2(I) is the adjustment to the positive OPT f2(I) limit required to compensate for each 1% that )Z,Y,X(FMQ exceeds BCDES(X,Y,Z) as defined in the COLR.
NSLOPEf2(I) is the adjustment to the negative OPT f2(I) limit required to compensate for each 1% that )Z,Y,X(FMQ exceeds BCDES(X,Y,Z) as defined in the COLR. * % f2(I) Margin is the most negative margin determined above.54 3.2.1 Insert Page B 3.2.1-5d INSERT 6 (continued) Completing Required Actions C.1 and C.2 is a conservative action for protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during an anticipated operational occurrence. Completing Required Actions C.1 and C.2 within the allowed Completion Time of 48 hours is sufficient considering the small likelihood of a limiting transient in this time period. 54 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-6 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES ACTIONS (continued) The implicit assumption is that if W(Z) values were recalculated (consistent with the reduced AFD limits), then )Z(FCQ times the recalculated W(Z) values would meet the FQ(Z) limit. Note that complying with this action (of reducing AFD limits) may also result in a power reduction. Hence the need for Required Actions B.2, B.3 and B.4.

B.2 A reduction of the Power Range Neutron Flux-High trip setpoints by 1% for each 1% by which the maximum allowable power is reduced, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER as a result of reducing AFD limits in accordance with Required Action B.1. B.3 Reduction in the Overpower T trip setpoints value of K4 by 1% for each 1% by which the maximum allowable power is reduced, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER as a result of reducing AFD limits in accordance with Required Action B.1. B.4 Verification that )Z(FWQ has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the maximum allowable power limit imposed by Required Action B.1 ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions. Condition B is modified by a Note that requires Required Action B.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action B.1, even when Condition A is exited prior to performing Required Action B.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER. 5 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-7 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES ACTIONS (continued) C.1 If Required Actions A.1 through A.4 or B.1 through B.4 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The Note applies REQUIREMENTS during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that )Z(FCQ and )Z(FWQ are within their specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which they were last verified to be within specified limits. Because )Z(FCQ and )Z(FWQ could not have previously been measured in this reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of )Z(FCQ and )Z(FWQ are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of )Z(FCQ and )Z(FWQ following a power increase of more than 10%, ensures that they are verified as soon as RTP (or any other level for extended operation) is achieved. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of )Z(FCQ and )Z(FWQ. The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured. D 551X,Y,Z A.5, B.1, B.2, C.1 or C.2 )Z,Y,X(FCQ)Z,Y,X(FMQ , SR 3.2.1.23)Z,Y,X(FCQ)Z,Y,X(FMQ )Z,Y,X(FMQ )Z,Y,X(FCQ)Z,Y,X(FMQ )Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FMQ 12 hours afterSurveillance performance is not required FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-8 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.2.1.1 Verification that )Z(FCQ is within its specified limits involves increasing)Z(FMQ to allow for manufacturing tolerance and measurement uncertainties in order to obtain )Z(FCQ. Specifically, )Z(FMQ is the measured value of FQ(Z) obtained from incore flux map results and )Z(FCQ = )Z(FMQ [1.0815] (Ref. 4). )Z(FCQ is then compared to its specified limits. The limit with which )Z(FCQ is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR. Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the )Z(FCQ limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased. If THERMAL POWER has been increased by 10% RTP since the last determination of )Z(FCQ, another evaluation of this factor is required [12] hours after achieving equilibrium conditions at this higher power level (to ensure that )Z(FCQ values are being reduced sufficiently with power increase to stay within the LCO limits).

[ The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS).

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] 1111267)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQDirect verification 3the overall measured FQ (X,Y,Z)INSERT 7 4 3.2.1 Insert Page B 3.2.1-8 INSERT 7 The surveillance has been modified by a Note providing an allowance to not perform SR 3.2.1.1 if the Surveillance has been determined to be met based on the performance results of both SR 3.2.1.2 and SR 3.2.1.3. If both the AFD min margin and the f2(I) min margin are positive, then the steady state limit is met because these margins represent bounding limiting conditions. However, if AFD min margin or f2(I) min margin is negative then a direct evaluation of the steady state limit is required to satisfy this surveillance requirement. 4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-9 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, )Z(FCQ, by W(Z) gives the maximum FQ(Z) calculated to occur in normal operation, )Z(FWQ. The limit with which )Z(FWQ is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR. The W(Z) curve is provided in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. )Z(FWQ evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive and b. Upper core region, from 85 to 100% inclusive.

The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions.

This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. If )Z(FWQ is evaluated, an evaluation of the expression below is required to account for any increase to )Z(FMQ that may occur and cause the FQ(Z) limit to be exceeded before the next required FQ(Z) evaluation. If the two most recent FQ(Z) evaluations show an increase in the expression maximum over z [ )Z(FCQ / K(Z) ], it is required to meet the FQ(Z) limit with the last )Z(FWQ increased by the greater of a factor of [1.02] or by an appropriate factor specified in the COLR (Ref. 5) and 3.2.1.3 5FQ (X,Y,Z)BQDES (X,Y,Z) and BCDES (X,Y,Z) limits INSERT 10 INSERT 8 INSERT 9 FQ (X,Y,Z))Z,Y,X(FMQ and found to be within the applicable limiting condition limits 244based on future projections 3.2.1 Insert Page B 3.2.1-9a INSERT 8 both assembly and axial location (X,Y,Z), has been included in the cycle specific limits BQDES(X,Y,Z) and BCDES(X,Y,Z) using margin factors MQ(X,Y,Z) and MC(X,Y,Z), respectively (Reference 5).

INSERT 9 No uncertainties are applied to )Z,Y,X(FMQ because the limits, BQDES(X,Y,Z) and BCDES(X,Y,Z), have been adjusted for uncertainties. 44 3.2.1 Insert Page B 3.2.1-9b INSERT 10 In addition to ensuring via surveillance that the heat flux hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in )Z,Y,X(FMQ for the last two measurements out to 31 EFPD beyond the most recent measurement. If the extrapolation yields an )Z,Y,X(FMQ > BQNOM(X,Y,Z), further consideration is required. The implications of these extrapolations are considered separately for both the operational and RPS heat flux hot channel factor limits. If the extrapolations of )Z,Y,X(FMQ are unfavorable, additional actions must be taken. These actions are to meet the FQ(X,Y,Z) limit with the last )Z,Y,X(FMQ increased by the appropriate factor specified in the COLR or to evaluate )Z,Y,X(FMQ prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements prevent FQ(X,Y,Z) from exceeding its limit for any significant period of time without detection using the best available data. Extrapolation is not required for the initial flux map taken after reaching equilibrium conditions following a refueling outage since the initial flux map establishes the baseline measurement for future trending. FQ(X,Y,Z) is verified at power levels 10% RTP above the THERMAL POWER of its last verification within 12 hours after achieving equilibrium conditions to ensure that FQ(X,Y,Z) is within its limit at higher power levels. 4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-10 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 1 Revision XXX 12BASES

SURVEILLANCE REQUIREMENTS (continued)

-----------------------------------REVIEWER'S NOTE----------------------------------- WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control and FQ Surveillance Technical Specification," February 1994, or other appropriate plant specific methodology, is to be listed in the COLR description in the Administrative Controls Section 5.0 to address the methodology used to derive this factor. -------------------------------------------------------------------------------------------------- or to evaluate FQ(Z) more frequently, each 7 EFPD. These alternative requirements prevent FQ(Z) from exceeding its limit for any significant period of time without detection. Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQ(Z) limit is met when RTP is achieved, because peaking factors are generally decreased as power level is increased. FQ(Z) is verified at power levels  10% RTP above the THERMAL POWER of its last verification, [12] hours after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.

The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of FQ(Z) evaluations. [ The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, when the plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day surveillances. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

-----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]  467 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B    WOG STS B 3.2.1B-11 Rev. 4.0,  1X,Y,Z212Revision XXX SEQUOYAH UNIT 1 BASES REFERENCES 1. 10 CFR 50.46, 1974.

2. Regulatory Guide 1.77, Rev. 0, May 1974.

3. 10 CFR 50, Appendix A, GDC 26.

4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988. 5. WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification," February 1994. 4BAW-10163PA "Core Operating Limit Methodology for Westinghouse-Designed PWRs" June 1989.

FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-12 Rev. 4.0, 1X,Y,Z212Revision XXX SEQUOYAH UNIT 1 4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-1 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12B 3.2 POWER DISTRIBUTION LIMITS

B 3.2.1B Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology) BASES BACKGROUND The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core. FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core. During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO(QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis. FQ(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution. FQ(Z) is measured periodically using the incore detector system. These measurements are generally taken with the core at or near equilibrium conditions.

Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ(Z) which are present during nonequilibrium situations such as load following or power ascension.

To account for these possible variations, the equilibrium value of FQ(Z) is adjusted as )Z(FWQby an elevation dependent factor that accounts for the calculated worst case transient conditions. Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. X,Y,Z 21X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z 1111and by assembly location, X, Y INSERT 1 8 3.2.1 Insert Page B 3.2.1-1 INSERT 1 "the FQ(X,Y,Z) limits, BQDES(X,Y,Z) and BCDES(X,Y,Z), have been adjusted by pre-calculated factors (MQ(X,Y,Z) and MC(X,Y,Z) respectively) to account for the calculated worst case transient conditions."1 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-2 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES

APPLICABLE This LCO precludes core power distributions that violate the following SAFETY fuel design criteria: ANALYSES a. During a large break loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200°F (Ref. 1), b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition,

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2), and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting. FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents

FQ(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships: FQ(Z) (CFQ / P) K(Z) for P > 0.5 FQ(Z) (CFQ / 0.5) K(Z) for P 0.5 where: CFQ is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) as a function of core height provided in the COLR, and P = THERMAL POWER / RTP 1X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z X,Y,Z RTPQFRTPQFFQ (X,Y,Z)X,Y,Z RTPQF1.8 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-3 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES

LCO (continued) For this facility, the actual values of CFQ and K(Z) are given in the COLR; however, CFQ is normally a number on the order of [2.32], and K(Z) is a function that looks like the one provided in Figure B 3.2.1B-1. For Relaxed Axial Offset Control operation, FQ(Z) is approximated by )Z(FCQ and )Z(FWQ. Thus, both )Z(FCQ and )Z(FWQ must meet the preceding limits on FQ(Z). An )Z(FCQ evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value ()Z(FMQ) of FQ(Z). Then, )Z(FCQ = )Z(FMQ [1.0815] where [1.0815] is a factor that accounts for fuel manufacturing tolerances and flux map measurement uncertainty. )Z(FCQ is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the incore flux map was taken. The expression for )Z(FWQ is: )Z(FWQ= )Z(FCQW(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR. The )Z(FCQ is calculated at equilibrium conditions. The FQ(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA. This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If )Z(FCQ cannot be maintained within the LCO limits, reduction of the core power is required and if )Z(FWQcannot be maintained within the LCO limits, reduction of the AFD limits is required. Note that sufficient reduction of the AFD limits will also result in a reduction of the core power. Violating the LCO limits for FQ(Z) produces unacceptable consequences if a design basis event occurs while FQ(Z) is outside its specified limits. RTPQFRTPQF2.62134INSERT 2 X,Y,Z )Z,Y,X(FCQX,Y,Z X,Y,Z 115SQN 3.2.1 Insert Page B 3.2.1-3a INSERT 2 Measured FQ(X,Y,Z) is compared against three limits:

Steady state limit, (FQRTP / P)
K(Z),
Limiting condition LOCA limit, BQDES(X,Y,Z), and
Limiting condition centerline fuel melt limit, BCDES(X,Y,Z). FQ(X,Y,Z) is approximated by )Z,Y,X(FCQ for the steady state limit on FQ(X,Y,Z). An )Z,Y,X(FCQ evaluation requires using the moveable incore detectors to obtain a power distribution map in MODE 1. From the incore flux map results we obtain the measured value ()Z,Y,X(FMQ) of FQ(X,Y,Z). Then, )Z,Y,X(FCQ = overall measured FQ(X,Y,Z)
1.05
1.03 where, 1.05 is the measurement reliability factor that accounts for flux map measurement uncertainty (Reference 5) and 1.03 is the local engineering heat flux hot channel factor to account for fuel rod manufacturing tolerance (Reference 4).

BQDES(X,Y,Z) and BCDES(X,Y,Z) are cycle dependent design limits to ensure the FQ(X,Y,Z) limit is met during transients. An evaluation of these limits requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the assembly nodal measured value ()Z,Y,X(FMQ) of FQ(X,Y,Z). )Z,Y,X(FMQ is compared directly to the limits BQDES(X,Y,Z) and BCDES(X,Y,Z). This is appropriate since BQDES(X,Y,Z) and BCDES(X,Y,Z) have been adjusted for uncertainties. The expression for BQDES(X,Y,Z) is: BQDES(X,Y,Z) = Pd(X,Y,Z)

MQ(X,Y,Z)
NRF
F1 / MRF where:
BQDES(X,Y,Z) is the cycle dependent maximum allowable design peaking factor for fuel assembly X,Y at axial location Z. BQDES(X,Y,Z) ensures that the LOCA limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties;
Pd(X,Y,Z) is the design power distribution for fuel assembly X,Y at axial location Z, including the operational flexibility factor;
MQ(X,Y,Z) is the minimum available margin ratio for the LOCA limit at assembly X,Y and axial location Z;
NRF is the nuclear reliability factor;
F1 is the spacer grid factor;
MRF is measurement reliability factor. 4 3.2.1 Insert Page B 3.2.1-3b INSERT 2 (continued) The expression for BCDES(X,Y,Z) is: BCDES(X,Y,Z) = Pd(X,Y,Z)
MC(X,Y,Z)
NRF
F1 / MRF where:
BCDES(X,Y,Z) is the cycle dependent maximum allowable design peaking factor for fuel assembly X,Y, at axial location Z. BCDES(X,Y,Z) ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties;
MC(X,Y,Z) is the minimum available margin ratio for the centerline fuel melt limit at assembly X,Y and axial location Z; The reactor core is operating as designed if the measured steady state core power distribution agrees with prediction within statistical variation. This guarantees that the operating limits will preserve the thermal criteria in the applicable safety analyses. The core is operating as designed if the following relationship is satisfied: )Z,Y,X(FMQ BQNOM(X,Y,Z) where:
BQNOM(X,Y,Z) is the nominal design peaking factor for assembly X,Y at axial location Z increased by an allowance for the expected deviation between the measured and predicted design power distribution. The FQ(X,Y,Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA. BQNOM (X,Y,Z), BQDES(X,Y,Z), and BCDES(X,Y,Z) Data bases are provided for the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in the COLR.

4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-4 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES

APPLICABILITY The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses. Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power. ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which )Z(FCQ exceeds its limit, maintains an acceptable absolute power density. )Z(FCQ is )Z(FMQmultiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. )Z(FMQ is the measured value of FQ(Z). The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of )Z(FCQ and would require power reductions within 15 minutes of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable power level. Decreases in )Z(FCQ would allow increasing the maximum allowable power level and increasing power up to this revised limit.

A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1% by which )Z(FCQ exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of )Z(FCQ and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. Decreases in )Z(FCQ would allow increasing the maximum allowable Power Range Neutron Flux - High trip setpoints. X,Y,Z 15)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FMQ 5)Z,Y,X(FMQ FQ (X,Y,Z))Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ4 )Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQINSERT 3 54 Move to next page after A.3 3.2.1 Insert Page B 3.2.1-4 INSERT 3 A.2 Required Action A.2 requires an administrative reduction of the AFD limits by 1% for each 1% by which )Z,Y,X(FCQ exceeds the steady state limit. The allowed Completion Time of 2 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factor limits are not exceeded. The maximum allowable AFD limits initially determined by Required Action A.2 may be affected by subsequent determinations of )Z,Y,X(FCQ and would require further AFD limit reductions within 2 hours of the )Z,Y,X(FCQdetermination, if necessary to comply with the decreased maximum allowable AFD limits. Decreases in),,(ZYXFCQwould allow increasing the maximum allowable AFD limits. 5 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-5 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES ACTIONS (continued) A.3 Reduction in the Overpower T trip setpoints (value of K4) by 1% for each 1% by which )Z(FCQ exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower T trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of )Z(FCQ and would require Overpower T trip setpoint reductions within 72 hours of the )Z(FCQ determination, if necessary to comply with the decreased maximum allowable Overpower T trip setpoints. Decreases in )Z(FCQ would allow increasing the maximum allowable Overpower T trip setpoints. A.4 Verification that )Z(FCQ has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions. Condition A is modified by a Note that requires Required Action A.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.

B.1 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, )Z(FWQ, exceeds its specified limits, there exists a potential for )Z(FCQ to become excessively high if a normal operational transient occurs. Reducing the AFD by 1% for each 1% by which )Z(FWQ exceeds its limit within the allowed Completion Time of 4 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factors are not exceeded. 51555154848)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ3 , SR 3.2.1.2 INSERT 4 INSERT 6 INSERT 5 in T span5543steady state and transient 55 , SR 3.2.1.2 , SR 3.2.1.2 35 )Z,Y,X(FCQ)Z,Y,X(FCQ 3.2.1 Insert Page B 3.2.1-5a INSERT 4 Since)Z,Y,X(FCQ exceeds the steady state limit, the limiting condition operational limit (BQDES) and the limiting condition Reactor Protection System limit (BCDES) may also be exceeded. By performing SR 3.2.1.2 and SR 3.2.1.3, appropriate actions with respect to reductions in AFD limits and OPT trip setpoints will be performed, ensuring that core conditions during operational and Condition II transients are maintained within the bounds of the safety analysis. 5 3.2.1 Insert Page B 3.2.1-5b INSERT 5 B.1 and B.2 The FQ(X,Y,Z) margin supporting AFD operational limits (AFD margin) during transient operations is based on the relationship between)Z,Y,X(FMQ and the limiting condition operational limit, BQDES (X,Y,Z), as follows:

%AFD margin = %*),,(),,(1001ZYXBQDESZYXFMQ The AFD min margin = minimum % margin value of all locations examined. If the reactor core is operating as designed, then )Z,Y,X(FMQ is less than BQDES (X,Y,Z) and calculation of %AFD margin is not required. If the AFD margin is less than zero, then )Z,Y,X(FMQ is greater than BQDES (X,Y,Z) and the AFD limits may not be adequate to prevent exceeding the peaking criteria for a LOCA if a normal operational transient occurs.

Required Actions B.1 and B.2 require reducing the AFD limit lines as follows. The AFD limit reduction is from the full power AFD limits. The adjusted AFD limits must be used until a new measurement shows that a smaller adjustment can be made to the AFD limits, or that no adjustment is necessary:

APL = PAFDL - Absolute Value of (PSLOPEAFD * % AFD Margin) ANL = NAFDL + Absolute Value of (NSLOPEAFD * % AFD Margin) where,

APL is the adjusted positive AFD limit.
ANL is the adjusted negative AFD limit.
PAFDL is the positive AFD limit defined in the COLR.
NAFDL is the negative AFD limit defined in the COLR.
PSLOPEAFD is the adjustment to the positive AFD limit required to compensate for each 1% that)Z,Y,X(FMQ exceeds BQDES (X,Y,Z) as defined in the COLR.
NSLOPEAFD is the adjustment to the negative AFD limit required to compensate for each 1% that)Z,Y,X(FMQ exceeds BQDES (X,Y,Z) as defined in the COLR. * % AFD Margin is the most negative margin determined above. Completing Required Actions B.1 and B.2 within the allowed Completion Time of 2 hours, restricts the axial flux distribution such that even if a transient occurred, core peaking factor limits are not exceeded. 54 3.2.1 Insert Page B 3.2.1-5c INSERT 6 C.1 and C.2 The FQ(X,Y,Z) margin supporting the Overpower T f2(I) breakpoints (f2(I) margin) during transient operations is based on the relationship between )Z,Y,X(FMQ and the limit, BCDES(X,Y,Z), as follows: % f2(I) margin = %*),,(),,(1001ZYXBCDESZYXFMQ The f2(I) min margin = minimum % margin value of all locations examined. If the reactor core is operating as designed, then )Z,Y,X(FMQ is less than BCDES(X,Y,Z) and calculation of % f2(I) margin is not required. If the f2(I) margin is less than zero, then )Z,Y,X(FMQ is greater than BCDES(X,Y,Z) and there is a potential that the f2(I) limits are insufficient to preclude centerline fuel melt during certain transients.

Required Actions C.1 and C.2 require reducing the f2(I) breakpoint limits as follows. The f2(I) breakpoint limit reduction is always from the full power f2(I) breakpoint limits. The adjusted f2(I) breakpoint limits must be used until a new measurement shows that a smaller adjustment can be made to the f2(I) breakpoint limits, or that no adjustment is necessary. Posf2(I)Adjusted = Posf2(I)Limit - Absolute Value of (PSLOPEf2(I) * % f2(I) Margin) Negf2(I)Adjusted = Negf2(I)Limit + Absolute Value of (NSLOPEf2(I) * % f2(I) Margin) where:

Posf2(I)Adjusted is the adjusted OPT positive f2(I) breakpoint limit.
Negf2(I)Adjusted is the adjusted OPT negative f2(I) breakpoint limit.
Posf2(I)Limit is the OPT positive f2(I) breakpoint limit defined in the COLR.
Negf2(I)Limit is the OPT negative f2(I) breakpoint limit defined in the COLR.
PSLOPEf2(I) is the adjustment to the positive OPT f2(I) limit required to compensate for each 1% that )Z,Y,X(FMQ exceeds BCDES(X,Y,Z) as defined in the COLR.
NSLOPEf2(I) is the adjustment to the negative OPT f2(I) limit required to compensate for each 1% that )Z,Y,X(FMQ exceeds BCDES(X,Y,Z) as defined in the COLR. * % f2(I) Margin is the most negative margin determined above.54 3.2.1 Insert Page B 3.2.1-5d INSERT 6 (continued) Completing Required Actions C.1 and C.2 is a conservative action for protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during an anticipated operational occurrence. Completing Required Actions C.1 and C.2 within the allowed Completion Time of 48 hours is sufficient considering the small likelihood of a limiting transient in this time period. 54 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-6 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES ACTIONS (continued) The implicit assumption is that if W(Z) values were recalculated (consistent with the reduced AFD limits), then )Z(FCQ times the recalculated W(Z) values would meet the FQ(Z) limit. Note that complying with this action (of reducing AFD limits) may also result in a power reduction. Hence the need for Required Actions B.2, B.3 and B.4.

B.2 A reduction of the Power Range Neutron Flux-High trip setpoints by 1% for each 1% by which the maximum allowable power is reduced, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER as a result of reducing AFD limits in accordance with Required Action B.1. B.3 Reduction in the Overpower T trip setpoints value of K4 by 1% for each 1% by which the maximum allowable power is reduced, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER as a result of reducing AFD limits in accordance with Required Action B.1. B.4 Verification that )Z(FWQ has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the maximum allowable power limit imposed by Required Action B.1 ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions. Condition B is modified by a Note that requires Required Action B.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action B.1, even when Condition A is exited prior to performing Required Action B.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER. 5 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-7 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES ACTIONS (continued) C.1 If Required Actions A.1 through A.4 or B.1 through B.4 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The Note applies REQUIREMENTS during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that )Z(FCQ and )Z(FWQ are within their specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which they were last verified to be within specified limits. Because )Z(FCQ and )Z(FWQ could not have previously been measured in this reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of )Z(FCQ and )Z(FWQ are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of )Z(FCQ and )Z(FWQ following a power increase of more than 10%, ensures that they are verified as soon as RTP (or any other level for extended operation) is achieved. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of )Z(FCQ and )Z(FWQ. The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured. D 551X,Y,Z A.5, B.1, B.2, C.1 or C.2 )Z,Y,X(FCQ)Z,Y,X(FMQ , SR 3.2.1.23)Z,Y,X(FCQ)Z,Y,X(FMQ )Z,Y,X(FMQ )Z,Y,X(FCQ)Z,Y,X(FMQ )Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FMQ 12 hours afterSurveillance performance is not required FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-8 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.2.1.1 Verification that )Z(FCQ is within its specified limits involves increasing)Z(FMQ to allow for manufacturing tolerance and measurement uncertainties in order to obtain )Z(FCQ. Specifically, )Z(FMQ is the measured value of FQ(Z) obtained from incore flux map results and )Z(FCQ = )Z(FMQ [1.0815] (Ref. 4). )Z(FCQ is then compared to its specified limits. The limit with which )Z(FCQ is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR. Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the )Z(FCQ limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased. If THERMAL POWER has been increased by 10% RTP since the last determination of )Z(FCQ, another evaluation of this factor is required [12] hours after achieving equilibrium conditions at this higher power level (to ensure that )Z(FCQ values are being reduced sufficiently with power increase to stay within the LCO limits).

[ The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS).

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] 1111267)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQ)Z,Y,X(FCQDirect verification 3the overall measured FQ (X,Y,Z)INSERT 7 4 3.2.1 Insert Page B 3.2.1-8 INSERT 7 The surveillance has been modified by a Note providing an allowance to not perform SR 3.2.1.1 if the Surveillance has been determined to be met based on the performance results of both SR 3.2.1.2 and SR 3.2.1.3. If both the AFD min margin and the f2(I) min margin are positive, then the steady state limit is met because these margins represent bounding limiting conditions. However, if AFD min margin or f2(I) min margin is negative then a direct evaluation of the steady state limit is required to satisfy this surveillance requirement. 4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-9 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, )Z(FCQ, by W(Z) gives the maximum FQ(Z) calculated to occur in normal operation, )Z(FWQ. The limit with which )Z(FWQ is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR. The W(Z) curve is provided in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. )Z(FWQ evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive and b. Upper core region, from 85 to 100% inclusive.

The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions.

This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. If )Z(FWQ is evaluated, an evaluation of the expression below is required to account for any increase to )Z(FMQ that may occur and cause the FQ(Z) limit to be exceeded before the next required FQ(Z) evaluation. If the two most recent FQ(Z) evaluations show an increase in the expression maximum over z [ )Z(FCQ / K(Z) ], it is required to meet the FQ(Z) limit with the last )Z(FWQ increased by the greater of a factor of [1.02] or by an appropriate factor specified in the COLR (Ref. 5) and 3.2.1.3 5FQ (X,Y,Z)BQDES (X,Y,Z) and BCDES (X,Y,Z) limits INSERT 10 INSERT 8 INSERT 9 FQ (X,Y,Z))Z,Y,X(FMQ and found to be within the applicable limiting condition limits 244based on future projections 3.2.1 Insert Page B 3.2.1-9a INSERT 8 both assembly and axial location (X,Y,Z), has been included in the cycle specific limits BQDES(X,Y,Z) and BCDES(X,Y,Z) using margin factors MQ(X,Y,Z) and MC(X,Y,Z), respectively (Reference 5).

INSERT 9 No uncertainties are applied to )Z,Y,X(FMQ because the limits, BQDES(X,Y,Z) and BCDES(X,Y,Z), have been adjusted for uncertainties. 44 3.2.1 Insert Page B 3.2.1-9b INSERT 10 In addition to ensuring via surveillance that the heat flux hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in )Z,Y,X(FMQ for the last two measurements out to 31 EFPD beyond the most recent measurement. If the extrapolation yields an )Z,Y,X(FMQ > BQNOM(X,Y,Z), further consideration is required. The implications of these extrapolations are considered separately for both the operational and RPS heat flux hot channel factor limits. If the extrapolations of )Z,Y,X(FMQ are unfavorable, additional actions must be taken. These actions are to meet the FQ(X,Y,Z) limit with the last )Z,Y,X(FMQ increased by the appropriate factor specified in the COLR or to evaluate )Z,Y,X(FMQ prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements prevent FQ(X,Y,Z) from exceeding its limit for any significant period of time without detection using the best available data. Extrapolation is not required for the initial flux map taken after reaching equilibrium conditions following a refueling outage since the initial flux map establishes the baseline measurement for future trending. FQ(X,Y,Z) is verified at power levels 10% RTP above the THERMAL POWER of its last verification within 12 hours after achieving equilibrium conditions to ensure that FQ(X,Y,Z) is within its limit at higher power levels. 4 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-10 Rev. 4.0, 1X,Y,Z2SEQUOYAH UNIT 2 Revision XXX 12BASES

SURVEILLANCE REQUIREMENTS (continued)

-----------------------------------REVIEWER'S NOTE----------------------------------- WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control and FQ Surveillance Technical Specification," February 1994, or other appropriate plant specific methodology, is to be listed in the COLR description in the Administrative Controls Section 5.0 to address the methodology used to derive this factor. -------------------------------------------------------------------------------------------------- or to evaluate FQ(Z) more frequently, each 7 EFPD. These alternative requirements prevent FQ(Z) from exceeding its limit for any significant period of time without detection. Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQ(Z) limit is met when RTP is achieved, because peaking factors are generally decreased as power level is increased. FQ(Z) is verified at power levels  10% RTP above the THERMAL POWER of its last verification, [12] hours after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.

The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of FQ(Z) evaluations. [ The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, when the plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day surveillances. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

-----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]  467 FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B    WOG STS B 3.2.1B-11 Rev. 4.0,  1X,Y,Z212Revision XXX SEQUOYAH UNIT 2 BASES REFERENCES 1. 10 CFR 50.46, 1974.

2. Regulatory Guide 1.77, Rev. 0, May 1974.

3. 10 CFR 50, Appendix A, GDC 26.

4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988. 5. WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification," February 1994. 4BAW-10163PA "Core Operating Limit Methodology for Westinghouse-Designed PWRs" June 1989.

FQ(Z) (RAOC-W(Z) Methodology) B 3.2.1B WOG STS B 3.2.1B-12 Rev. 4.0, 1X,Y,Z212Revision XXX SEQUOYAH UNIT 2 4 JUSTIFICATION FOR DEVIATIONS ITS 3.2.1, BASES, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The RAOC-W(Z) methodology and the Specification designator "B" are deleted because they are unnecessary. (Only one Heat Flux Hot Channel Factor Specification is used in the SQN ITS.) This information is provided in NUREG-1431, Rev. 4 to assist in identifying the appropriate Specification to be used as a model for the plant specific ITS conversion, but serves no purpose in a plant specific implementation. In addition, the CAOC-FXY and CAOC-W(Z) methodology Specification Bases (ISTS B 3.2.1A and B 3.2.1C) are not used and are not shown.

3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is changed to reflect the current licensing basis.

4. The ISTS Bases for LCO 3.2.1, has been updated to reflect the methodology identified in BAW-10163PA "Core Operating Limit Methodology for Westinghouse-Designed PWRs" June 1989.

5. Changes have been made to be consistent with changes made to the Specification. 6. ISTS SR 3.2.1.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies for ITS SR 3.2.1.1 under the Surveillance Frequency Control Program. 7. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

8. Editorial changes made to enhance clarity/consistency.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.2.1, HEAT FLUX HOT CHANNEL FACTOR (FQ(X,Y,Z)) Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 2 ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (FH(X,Y)) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) LIMITING CONDITION FOR OPERATION 3.2.3 FH(X,Y) shall be maintained within the limits specified in the COLR. APPLICABILITY: MODE 1 ACTION: With FH(X,Y) exceeding the limit specified in the COLR:

a. Within 2 hours either:

1. Restore FH(X,Y) to within the limit specified in the COLR, or 2. Reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH*% for each 1% that FH(X,Y) exceeds the limit, and

b. Within the next 4 hours either:

1. Restore FH(X,Y) to within the limit specified in the COLR, or 2. Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH*% for each 1% that FH(X,Y) exceeds that limit, and

c. Within 24 hours of initially being outside the limit specified in the COLR, either: 1. Restore FH(X,Y) to within the limit specified in the COLR, or 2. Verify through incore flux mapping that FH(X,Y) is restored to within the limit for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.

_______________________

RRH is the amount of power reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19, 138, 155, 223 LCO 3.2.2 Applicability ACTION A Required Action A.2 LA01Required Action A.3 ACTION A ACTION A Required Action A.1 Page 1 of 8 Add proposed ACTION A NoteM01LA01LA01LA02L0172M02Add proposed ACTION C A02A02A02ACTION C L026 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS ACTION: (Continued)

d. Within 48 hours of initially being outside the limit specified in the COLR, reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH** for each 1% that FH(X,Y) exceeds the limit, and

e. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b. and/or c. and/or d., above: subsequent POWER OPERATION may proceed provided that FH(X,Y) is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER, 2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.

___________________________ ** TRH is the amount of Overtemperature Delta T K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-11 Amendment No. 138, 223 Required Action A.4 Required Action A.5 Completion Time A.5 LA03Page 2 of 8 Add proposed Required Action A.5 Note LA02LA03A03 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 HMF(X,Y) shall be evaluated to determine if FH(X,Y) is within its limit by: a. Using the movable incore detectors to obtain a power distribution map HMF(X,Y)* at any THERMAL POWER greater than 5% of RATED THERMAL POWER. b. Satisfying the following relationship: FHRM(X,Y) BHNOM(X,Y) Where: And BHNOM(X,Y)** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement. MAPM is the maximum Allowable Peak** obtained from the measured power distribution. AXIAL(X,Y) is the axial shape for FH(X,Y).

c. If the above relationship is not satisfied, then 1. For the location, calculate the % margin to the maximum allowable design as follows: 100% x Y)BRDES(X,Y)(X,HRF 1 = I)Margin(f %M1 where BHDES (X,Y) and BRDES (X,Y)** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within the LCO limits, and include allowances for calculational and measurement uncertainties. _________________________

No additional uncertainties are required in the following equations for HMF(X,Y) and F)HRM(X,Y), because the limits include uncertainties. ** BHNOM(X,Y), MAPM, BHDES(X,Y), and BRDES(X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14. April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-11a Amendment No. 223 FHR(X,Y) = F(X,Y)MAP / AXIAL(X,Y)MHMM % F Margin = 1 FHR(X,Y)BHDES(X,Y) x 100%HM M03LA04Page 3 of 8 SR 3.2.2.1 SR 3.2.2.2 SR NOTE LA04SR 3.2.2.2 SR 3.2.2.1 LA04 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 2. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above. FH min margin = minimum % margin value of all locations examined f1(I) min margin = minimum % margin value of all locations examined 3. If the FH min margin in 4.2.3.2.c.2 above is < 0, then within 2 hours reduce the allowable THERMAL POWER from RATED THERMAL POWER by RRH*% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply.

4. If the f1(I) min margin in 4.2.3.2.c.2 above is < 0, then within 48 hours reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH**% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the action statements for 3.2.3 apply. d. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding FHRM(X,Y) > BHNOM(X,Y) either of the following actions shall be taken:

1. HMF(X,Y) shall be increased over that specified in 4.2.3.2.a by the appropriate factor specified in the COLR, and 4.2.3.2.c.1 repeated, or

2. HMF(X,Y) shall be evaluated according to 4.2.3.2 at or before the time when the margin is projected to result in the action specified in 4.2.3.2.c.3 or 4.2.3.2.c.4. 4.2.3.3 Y)(X,FMH shall be determined to be within its limit by using the incore detectors to obtain a power distribution map: a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 EFPD. ____________________________

RRH is the amount of power reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.
- TRH is the amount of Overtemperature Delta T K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-11b Amendment No. 223 LA01SR 3.2.2.1 SR 3.2.2.2 Page 4 of 8 A04thereafter LA05In accordance with the Surveillance Frequency Control Program LA04ACTION A ACTION B SR 3.2.2.1/SR 3.2.2.2 NOTE SR 3.2.2.1/SR 3.2.2.2 NOTE a. SR 3.2.2.1/SR 3.2.2.2 NOTE b. SR 3.2.2.1 SR 3.2.2.2 LA02M04M04LA03LA01LA03 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) LIMITING CONDITION FOR OPERATION 3.2.3 FH(X,Y) shall be maintained within the limits specified in the COLR. APPLICABILITY: MODE 1 ACTION: With FH(X,Y) exceeding the limit specified in the COLR: a. Within 2 hours either:

1. Restore FH(X,Y) to within the limit specified in the COLR, or 2. Reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH*% for each 1% that FH(X,Y) exceeds the limit, and b. Within the next 4 hours either: 1. Restore FH(X,Y) to within the limit specified in the COLR, or

2. Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH*% for each 1% that FH(X,Y) exceeds that limit, and

c. Within 24 hours of initially being outside the limit specified in the COLR, either:

1. Restore FH(X,Y) to within the limit specified in the COLR, or

2. Verify through incore flux mapping that FH(X,Y) is restored to within the limit for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours.

RRH is the amount of power reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-8 Amendment No. 21, 130, 146, 214 LCO 3.2.2 Applicability ACTION A Required Action A.2 LA01Required Action A.3 ACTION A ACTION A Required Action A.1 Page 5 of 8 Add proposed ACTION A NoteM01LA01LA01LA02L0172M02Add proposed ACTION C A02A02A02ACTION C L026 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS ACTION: (Continued) d. Within 48 hours of initially being outside the limit specified in the COLR, reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH** for each 1% that FH(X,Y) exceeds the limit, and

e. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b. and/or c. and/or d., above; subsequent POWER OPERATION may proceed provided that FH(X,Y) is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels: 1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.

     ** TRH is the amount of Overtemperature Delta T K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-9 Amendment No. 130, 214 Required Action A.4 Required Action A.5 Completion Time A.5 LA03Page 6 of 8 LA03LA02A03Add proposed Required Action A.5 Note A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 HMF(X,Y) shall be evaluated to determine if FH(X,Y) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map HMF(X,Y)* at any THERMAL POWER greater than 5% of RATED THERMAL POWER. b. Satisfying the following relationship: FHRM(X,Y) BHNOM (X,Y)

Where: And BHNOM (X,Y)** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement. MAPM is the maximum Allowable Peak** obtained from the measured power distribution. AXIAL (X,Y) is the axial shape for FH(X,Y).

c. If the above relationship is not satisfied, then

1. For the location, calculate the % margin to the maximum allowable design as follows: where BHDES(X,Y) and BRDES(X,Y)** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within the LCO limits, and include allowances for calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for HMF(X,Y)1 and FHRM(X,Y), because the limits include uncertainties. ** BHNOM (X,Y), MAPM, BHDES (X,Y), and BRDES (X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-9a Amendment No. 214 Y)AXIAL(X, / MAPY)(X,F = Y)(X,HRFMMHM 100% x Y)BRDES(X,Y)(X,HRF 1 = I)Margin(f %M1 100% x BHDES(X,Y)(X,Y)HRF 1 = MarginF %MH M03Page 7 of 8 SR 3.2.2.1 SR 3.2.2.2 SR 3.2.2.2 SR 3.2.2.1 SR NOTE LA04LA04LA04 A01ITS 3.2.2 ITS POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 2. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above. FH min margin = minimum % margin value of all locations examined f1(I) min margin = minimum % margin value of all locations examined

3. If the FH min margin in 4.2.3.2.c.2 above is < 0, then within 2 hours reduce the allowable THERMAL POWER from RATED THERMAL POWER by RRH*% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply.

4. If the f1(I) min margin in 4.2.3.2.c.2 above is < 0, then within 48 hours reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH**% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the action statements for 3.2.3 apply. d. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding FHRM (X,Y) > BHNOM (X,Y) either of the following actions shall be taken:

1. Y)(X,FMH shall be increased over that specified in 4.2.3.2.a by the appropriate factor specified in the COLR, and 4.2.3.2.c.1 repeated, or 2. HMF(X,Y) shall be evaluated according to 4.2.3.2 at or before the time when the margin is projected to result in the action specified in 4.2.3.2.c.3 or 4.2.3.2.c.4. 4.2.3.3 HMF(X,Y) shall be determined to be within its limit by using the incore detectors to obtain a power distribution map: a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b. At least once per 31 EFPD.

RRH is the amount of power reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14. ** TRH is the amount of Overtemperature Delta T K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-9b Amendment No. 214 LA01SR 3.2.2.1 SR 3.2.2.2 Page 8 of 8 LA04M04LA02LA05In accordance with the Surveillance Frequency Control Program A04thereafter SR 3.2.2.1 SR 3.2.2.2 ACTION A ACTION B SR 3.2.2.1/SR 3.2.2.2 NOTE SR 3.2.2.1/SR 3.2.2.2 NOTE a. SR 3.2.2.1/SR 3.2.2.2 NOTE b M04LA03LA01LA03 DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 1 of 8 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.2.3 ACTION a.1, b.1 and c.1 require the restoration of FH(X,Y) to within the limit specified in the COLR. ISTS LCO 3.0.2 Bases states that correction of the entered Condition is an action that may always be considered upon entering ACTIONS and that the restoration of compliance with the LCO is always an option. This changes the CTS by not specifically stating that restoration of FH(X,Y) is required. This change is acceptable because the technical requirements have not changed. ISTS LCO 3.0.2 Bases states that correction of the entered Condition is an action that may always be considered upon entering ACTIONS and that the restoration of compliance with the LCO is always an available Required Action. The convention in the ITS is to not state such "restore" options explicitly unless it is the only action or is required for clarity. In this specific application, Required Action A.1.1 is not the only ACTION and a power reduction should be the focus for restoration of FH(X,Y) to within the limits. This change is designated as administrative, because it does not result in technical changes to the CTS. A03 CTS 3.2.3 ACTION e states in part that with FH(X,Y) exceeding its limit, FH(X,Y) must be demonstrated to be within its limit prior to exceeding 50% RTP and 75% RTP, and within 24 hours of attaining or exceeding 95% RTP. ITS 3.2.2 Required Action A.5 contains the same requirements. However, ITS 3.2.2 Required Action A.5 is modified by a Note which states "THERMAL POWER does not have to be reduced to comply with this Required Action." This modifies the CTS by adding a Note stating that THERMAL POWER does not have to be reduced to comply with the Required Action. This change is acceptable, because the requirements have not changed. The Note is included in the ITS to make clear that THERMAL POWER does not have to be reduced to perform the Required Action. For example, if FH(X,Y) exceeds its limit and, per ITS Required Action A.1, THERMAL POWER is reduced to 60% RTP, THERMAL POWER does not have to be reduced to less than 50% RTP to verify FH(X,Y) is within its limit to comply with ITS Required Action A.5. However, FH(X,Y) must still be measured prior to exceeding 75% RTP and within 24 hours of attaining or exceeding 95% RTP. The Note is needed because the ITS contains a Note in ITS 3.2.2 ACTION A that states "Required Actions A.3 and A.5 must be completed whenever Condition A is entered." The ITS 3.2.2 ACTION A Note does not exist in the CTS and could be construed as DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 2 of 8 requiring THERMAL POWER to be reduced to comply with Required Action A.5. (Addition of the ACTION A Note is discussed in DOC M01.) As a result, the Required Action A.5 Note makes the ITS and CTS actions consistent. This change is designated as administrative, because it does not result in technical changes to the CTS. A04 CTS 4.2.3.3 requires Y)(X,FMHto be determined prior to operation above 75% of RTP after each fuel loading, and at least once per 31 EFPD. ITS SR 3.2.2.1 and SR 3.2.2.2 Frequency is once after each refueling prior to THERMAL POWER exceeding 75% RTP AND 31 EFPD thereafter. This changes the CTS by adding the word "thereafter" to the Frequency. The removal of the "31 EFPD thereafter" Frequency to the Surveillance Frequency Control Program is discussed in DOC LA05. CTS 4.2.3.3 is required to be performed prior to operation above 75% RTP after each fuel loading and once per 31 EFPD. Also, although this Frequency is removed to the Surveillance Frequency Control Program, the addition of the word "thereafter" in ITS SR 3.2.2.1 and SR 3.2.2.2 ensures that the 31 EFPD Frequency starts after the first performance of the SR, which is required prior to exceeding 75% RTP after each fuel loading. Therefore, the addition of the word "thereafter" is considered acceptable because the use of thereafter is essentially the same as the CTS Frequency. This change is designated as administrative, because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 3.2.3 ACTION e states that with FH(X,Y) exceeding its limit "subsequent POWER OPERATION may proceed provided that FH(X,Y) is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels: 1. A nominal 50% of RATED THERMAL POWER, 2. A nominal 75% of RATED THERMAL POWER, and 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER." However, under CTS 3.0.2, these measurements do not have to be completed, if compliance with the LCO is restored. ITS 3.2.2 ACTION A contains a Note which states, "Required Actions A.3 and A.5 must be completed whenever Condition A is entered." ITS 3.2.2 Required Action A.3 requires verification that FH min margin is >0 24 hours after entry into Condition A. Required Action A.5 requires verification that FH min margin is >0 prior to THERMAL POWER exceeding 50% RTP and 75% RTP, and within 24 hours after THERMAL POWER is greater than or equal to 95% RTP. This changes the CTS by requiring the verification that FH min margin is >0 to be made even if FH(X,Y) is restored to within its limit. This change is acceptable, because it establishes appropriate compensatory measurements for violation of the FH(X,Y) limit. As power is reduced under ITS 3.2.2 Required Action A.1, the margin to the FH(X,Y) limit increases. Therefore, compliance with the LCO could be restored during the power reduction. Verifying that the limit is met as power is increased ensures that the limit continues to be met and does not remain unmeasured for up to 31 EFPD. This change is DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 3 of 8 designated as a more restrictive change because it imposes requirements in addition to those in the CTS. M02 CTS 3.2.3 does not contain an Action to follow if ACTIONS a, b, d, and e cannot be met. Therefore, CTS 3.0.3 would be entered, which would allow 1 hour to initiate a shutdown and to be in HOT STANDBY within 7 hours. ITS 3.2.2 ACTION C, states that the plant must be in MODE 2 within 6 hours, if any Required Action and associated Completion Time is not met. This changes the CTS by eliminating the one hour to initiate a shut down and, consequently, allowing one hour less for the unit to be in MODE 2. The purpose of CTS 3.0.3 is to delineate the ACTION to be taken for circumstances not directly provided for in the ACTION statement and whose occurrences would violate the intent of the Specification. This change is acceptable because it provides an appropriate compensatory measure for the described conditions. If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. The LCO is applicable in MODE 1. Requiring a shut down to MODE 2 is appropriate in this condition. The one hour allowed by CTS 3.0.3 to prepare for a shut down is not needed, because the operators have had time to prepare for the shut down while attempting to follow the Required Actions and associated Completion Times. This change is designated as more restrictive because it allows less time to shut down than does the CTS. M03 CTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable provides an allowance for entering the next higher MODE of Applicability when the LCO is not met. ITS 3.2.2 has no specific allowance for changing MODES at any time with ITS LCO 3.2.2 not met. ITS LCO 3.0.4 requires in part that, When an LCO is not met, entry into a MODE or other specified Condition in the Applicability shall only be made: If either part a. or part b. or part c. is met. Part c provides the following allowance, When an allowance is stated in the individual value, parameter or other specification. ITS 3.2.2 Surveillance Requirements Note will be added to provide the following allowance, Not required to be performed until 12 hours after an equilibrium power level has been achieved, at which a power distribution map can be obtained. This changes CTS by allowing entry into the MODE of Applicability by only deferring the performance of the Surveillance Requirements instead of deferring compliance with the LCO. The purpose of CTS 4.2.3.1 is to provide an exception to SR 4.0.4. SR 4.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability. This change is acceptable because ITS provides an allowance to enter the MODE of Applicability at any time ITS LCO 3.2.2 is not met solely based on surveillance performance. SR 3.2.2.1 and SR 3.2.2.2 require using the incore detector system to provide the necessary data to create a power distribution map. To provide the necessary data, MODE 1 needs to be entered, power escalated, stabilized and equilibrium conditions established at some higher power level (~40%-50%). The surveillances cannot be performed in MODE 2. This change is designated as more restrictive because the CTS 4.0.4 MODE change allowance for not met is now limited to DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 4 of 8 the performance of the SRs and does not include the allowance to change MODES with the acceptance criteria not met. M04 CTS 3.2.3 provides two acceptable alternatives for the FH min margin and f1(I) min margin not met. CTS 4.2.3.2.c.3 states, If the FH min margin in 4.2.3.2.c.2 above is < 0, then within 2 hours reduce the allowable THERMAL POWER from RATED THERMAL POWER by RRH*% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply. CTS 4.2.3.2.c.4 states, If the f1(I) min margin in 4.2.3.2.c.2 above is < 0, then within 48 hours reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH**% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the action statements for 3.2.3 apply. CTS 4.2.3.2.c.3 has been replaced by ITS 3.2.2 Required Actions A.1. Similarly, CTS 4.2.3.2.c.4 has been replaced with ITS 3.2.2 Required Actions B.1. However, in both cases the option for, otherwise, the action statements for 3.2.3 apply has not been retained. This changes the CTS by removing the option to follow the action statement of CTS 3.2.3 for either min margin (FH or f1(I)) not met. The purpose of CTS 4.2.3.2.c.3 and CTS 4.2.3.2.c.4 is to provide acceptable alternatives for the required compensatory actions when either FH min margin or f1(I) min margin is not met. The CTS surveillance requirements for FH min margin not met requires the reduction of ALLOWABLE THERMAL POWER from RTP by RRH*% x most negative margin from 4.2.3.2.c.2. This requirement is being retained as ITS 3.2.2 Required Action A.1. The CTS surveillance requirements for f1(I) min margin not met requires the reduction of the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH**% x most negative margin from 4.2.3.2.c.2. This requirement is being retained as ITS 3.2.2 Required Action B.1. If the ITS Required Actions are not performed within the allowed Completion Time, Condition C will be entered requiring the Unit to be placed in MODE 2. This change is designated as more restrictive because an acceptable alternative Required Action available in CTS is being removed. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.2.3 provides actions to take within 2 hours when FH(X,Y) is not within limits, and states to reduce the allowable THERMAL POWER and within 4 hours reduce the Power Range Neutron Flux-High Trip Setpoint at least RRH*% for each 1% that FH(X,Y) exceeds the limit provided in the COLR. Similarly, CTS 4.2.3.2.c.3 requires in part to reduce the allowable THERMAL POWER from RATED THERMAL POWER by RRH*% x most negative margin from 4.2.3.2.c.2. CTS NOTE

provides the definition of RRH as the amount of power reduction required to compensate for each 1% that FH(X,Y)

DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 5 of 8 exceeds the limit provided in the COLR per Specification 6.9.1.14. ITS 3.2.2 Required Action A.1 requires within 2 hours of discovery that FH min margin is not within limits, to reduce THERMAL POWER from RTP, and ITS 3.2.2 Required Action A.2 requires within 72 hours to reduce the Power Range Neutron Flux-High Trip Setpoint by RRH% multiplied times the FH min margin. This changes the CTS by relocating the definition of RRH to the COLR. The removal of these details from the Technical Specifications and its relocation into the COLR is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements to reduce THERMAL POWER from RTP and reduce the Power Range Neutron Flux-High Trip Setpoint by RRH% for each 1% that FH(X,Y) exceeds its limit. The definition of RRH is already located in the COLR. Also, this change is acceptable because the removed information will be adequately controlled in the COLR requirements provided in ITS 5.6.5, "Core Operating Limits Report." ITS 5.6.5 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such transient analysis limits and accident analysis limits) of the safety analyses are met. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.2.3 ACTIONS c.2 and e require FH(X,Y) to be determined to be within its limit through incore flux mapping. Additionally, CTS 4.2.3.3 requires HMF(X,Y)to be determined to be within its limit by using the incore detectors to obtain a power distribution map. ITS SR 3.2.2.1 and SR 3.2.2.2 collectively verifiy that FH(X,Y) is within its limit. This changes the CTS by moving the manner in which the FH(X,Y) determination is performed to the Bases. The removal of these details for performing actions and a Surveillance Requirement from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to determine FH(X,Y) is within its limit. Also, this change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.2.3 Action d requires within 48 hours of FH(X,Y) being outside its limits, to reduce the Overtemperature Delta T K1 term in Table 2.2-1 by at least TRH** for each 1% that FH(X,Y) exceeds the limit. Similarly, CTS 4.2.3.2.c.4 requires in part to reduce Overtemperature Delta T K1 term in DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 6 of 8 Table 2.2-1 by at least TRH** x most negative margin from 4.2.3.2.c.2. CTS Note ** provides a definition for TRH as the amount of Overtemperature Delta T K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR. ITS 3.2.2 Required Action A.4 states when FH min margin is < 0, reduce the OTT setpoint by TRH multiplied times the f1(I) min margin. This changes the CTS by moving the details of the specific variable within OTT to be reduced, the location of the K1 terms, and the definition of TRH to the COLR. The removal of these details from the Technical Specifications and their relocation into the COLR is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to reduce the OTT setpoint by TRH multiplied times the f1(I) min margin. The specific variable within OTT to be reduced, the location of the K1 terms, and definition of TRH are already located in the COLR. Also, this change is acceptable because the removed information will be adequately controlled in the COLR requirements provided in ITS 5.6.5, "Core Operating Limits Report." ITS 5.6.5 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such transient analysis limits and accident analysis limits) of the safety analyses are met. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA04 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.2.3.2.a, 4.2.3.2.b, 4.2.3.2.c.1, and 4.2.3.2.c.2, provide details for evaluating FMH(X,Y) to determine if FH(X,Y) is within limits. ITS SR 3.2.2.1 and SR 3.2.2.2 collectively verify that FH(X,Y) is within limits specified in the COLR. This changes the CTS by moving the details for evaluating FMH(X,Y) to determine if FH(X,Y) is within limits to the TS Bases. The removal of these details from the Technical Specifications and their relocation into the TS Bases is acceptable, because the procedural steps and further details for making a determination that FH(X,Y) is within its limits is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to determine FH(X,Y) is within its limits specified in the COLR. Also, this change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA05 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.2.3.3 requires, in part, a determination that FH(X,Y) is within its limits at least once per 31 EFPD. ITS SR 3.2.2.1 and SR 3.2.2.2 collectively require a similar Surveillance and specify the periodic Frequency as, "In DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 7 of 8 accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequency for this SR and associated Bases to the Surveillance Frequency Control Program The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.2.3 ACTION b states, in part, that when FH(X,Y) exceeds its limit, reduce the Power Range Neutron Flux - High Trip setpoints by at least RRH*% for each 1% FH(X,Y) exceeds that limit within the next 4 hours. ITS 3.2.2 Required Actions A.2 states with FH(X,Y) not within limit, reduce the Power Range Neutron Flux - High trip setpoints by at least RRH% multiplied times the FH min margin within 72 hours. This changes the CTS by increasing the time allowed to reduce the trip setpoints. The purpose of CTS 3.2.3 ACTION b is to lower the Power Range Neutron Flux - High Trip setpoints, which ensures continued operation is at an acceptably low power level with an adequate DNBR margin and avoids violating the FH(X,Y) limit. This change is acceptable, because the Completion Time is consistent with safe operation and recognizes that the safety analysis assumptions are satisfied once power is reduced, and considers the low probability of a DBA occurring during the allowed Completion Time. The revised Completion Time allows the Power Range Neutron Flux - High Trip setpoints to be reduced in a controlled manner without challenging operators, technicians, or plant systems. Following a significant power reduction, a time period of 24 hours is allowed to reestablish steady state xenon concentration and power distribution and to take and analyze a flux map. If it is determined that FH(X,Y) is still not within its limit, reducing the Power Range Neutron Flux - High Trip Setpoints can be accomplished within a few hours. Furthermore, setpoint changes should only be required for extended operation in this condition, because of the risk of a plant trip during the adjustment. This change is designated as less restrictive, because additional time is allowed to lower the Power Range Neutron Flux - High Trip setpoints than was allowed in the CTS. DISCUSSION OF CHANGES ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FH(X,Y) Sequoyah Unit 1 and Unit 2 Page 8 of 8 L02 (Category 3 - Relaxation of Completion Time) CTS 3.2.3 ACTION c.2 states, "Verify through incore flux mapping that FH(X,Y) is restored to within the limit for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next two hours." ITS 3.2.2 ACTION C states, "Required Action and associated Completion Time not met." Required Action C.1 states, "Be in MODE 2" within a Completion Time of "6 hours." This changes the CTS by increasing the time allowed to exit the MODE of Applicability when the Required Actions or associated Completion Times are not met. The purpose of CTS 3.2.3 ACTION c.2 is to, within 24 hours, either verify FH(X,Y) is restored within limits for the reduced power level or within the next 2 hours, enter MODE 2. Under similar conditions, ITS will require the plant to be placed in a MODE in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems. This change is acceptable, because the Completion Time is consistent with safe operation and recognizes that the safety analysis assumptions are satisfied once power is reduced. This change is designated as less restrictive, because additional time is allowed to exit the LCO than was allowed in the CTS. . Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) NHF 3.2.2 WOG STS 3.2.2-1 Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 1 Amendment xxx 113.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (NHF) LCO 3.2.2 NHF shall be within the limits specified in the COLR. APPLICABILITY: MODE 1. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ Required Actions A.2 and A.3 must be completed whenever Condition A is entered. --------------------------------- NHFnot within limit. A.1.1 Restore NHF to within limit. OR A.1.2.1 Reduce THERMAL POWER to < 50% RTP. AND A.1.2.2 Reduce Power Range Neutron Flux - High trip setpoints to 55% RTP. AND A.2 Perform SR 3.2.2.1 AND 4 hours

4 hours 72 hours

24 hours FH(X,Y)FH(X,Y) INSERT 1 INSERT 333.2.3 Applicability DOC M01

ACTION a.2 SR 4.2.3.2.c.3 ACTION b.2 ACTION c.2 112435653INSERT 2 FH min margin < 0 2 5 3 SR 4.2.3.2.c.3 222allowable 3.2.2 Insert Page 3.2.2-1 CTS INSERT 1 from RTP by RRH% multiplied times the FH min margin.

INSERT 2 by RRH% multiplied times the FH min margin.

INSERT 3 A.4 Reduce Overtemperature T trip setpoint by TRH multiplied times the FH min margin. AND 48 hours ACTION d 355ACTION a.2 4.2.3.2.c.3 ACTION b.2 NHF 3.2.2 WOG STS 3.2.2-2 Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 1 Amendment xxx 11ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.3 --------------NOTE-------------- THERMAL POWER does not have to be reduced to comply with this Required Action. ------------------------------------- Perform SR 3.2.2.1.

Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours after THERMAL POWER reaching 95% RTP B. Required Action and associated Completion Time not met. B.1 Be in MODE 2. 6 hours 5ACTION e DOC A03 ACTION e.1 ACTION e.2 ACTION e.3 ACTION c.2 DOC M02 3C CINSERT 4 66 3.2.2 Insert Page 3.2.2-2 CTS INSERT 4 CONDITION REQUIRED ACTION COMPLETION TIME B. f1(I) min margin < 0. B.1 Reduce Overtemperature T trip setpoint by TRH multiplied times the f1(I) min margin. 48 hours 4.2.3.2.c.4 6 NHF3.2.2 WOG STS3.2.2-3Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 1Amendment xxx 11SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify NHF is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 8INSERT 7INSERT 587INSERT 64.2.3.2.c.1 4.2.3.3.a 4.2.3.3.b 899FH min margin > 0 3.2.2 Insert Page 3.2.2-3a CTS INSERT 5 ------------------------------------------------------------NOTE----------------------------------------------------------- Not required to be performed until 12 hours after an equilibrium power level has been achieved, at which a power distribution map can be obtained.

74.2.3.1 3.2.2 Insert Page 3.2.2-3b CTS INSERT 6 -------------------------------NOTE------------------------------If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: FHRM(X,Y) > BHNOM(X,Y) a.Increase HMF(X,Y) by the appropriate factorspecified in the COLR and reverifyFH min margin > 0; orb.Repeat SR 3.2.2.1 prior to the time at which theprojected FH min margin will be < 0.--------------------------------------------------------------------- 4.2.3.2.d 4.2.3.2.d.1 4.2.3.2.d.2 3.2.2 Insert Page 3.2.2-3c CTS INSERT 7 SURVEILLANCE FREQUENCY SR 3.2.2.2 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: FHRM(X,Y) > BHNOM(X,Y) a. Increase HMF(X,Y) by the appropriate factorspecified in the COLR and reverifyf1 (I) min margin > 0; orb.Repeat SR 3.2.2.2 prior to the time at whichthe projected f1 (I) min margin will be < 0.--------------------------------------------------------------------- Verify f1(I) min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND In accordance with the Surveillance Frequency Control Program 84.2.3.2.d 4.2.3.2.c.1 4.2.3.3.a 4.2.3.3.b 4.2.3.2.d.1 4.2.3.2.d.2 NHF 3.2.2 WOG STS 3.2.2-1 Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 2 Amendment xxx 113.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (NHF) LCO 3.2.2 NHF shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ Required Actions A.2 and A.3 must be completed whenever Condition A is entered. --------------------------------- NHFnot within limit. A.1.1 Restore NHF to within limit. OR A.1.2.1 Reduce THERMAL POWER to < 50% RTP. AND A.1.2.2 Reduce Power Range Neutron Flux - High trip setpoints to 55% RTP. AND A.2 Perform SR 3.2.2.1 AND 4 hours

4 hours 72 hours

24 hours FH(X,Y)FH(X,Y) INSERT 1 INSERT 333.2.3 Applicability DOC M01

ACTION a.2 SR 4.2.3.2.c.3 ACTION b.2 ACTION c.2 112435653INSERT 2 FH min margin < 0 2 5 3 SR 4.2.3.2.c.3 222allowable 3.2.2 Insert Page 3.2.2-1 CTS INSERT 1 from RTP by RRH% multiplied times the FH min margin.

INSERT 2 by RRH% multiplied times the FH min margin.

INSERT 3 A.4 Reduce Overtemperature T trip setpoint by TRH multiplied times the FH min margin. AND 48 hours ACTION d 355ACTION a.2 4.2.3.2.c.3 ACTION b.2 NHF 3.2.2 WOG STS 3.2.2-2 Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 2 Amendment xxx 11ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.3 --------------NOTE-------------- THERMAL POWER does not have to be reduced to comply with this Required Action. ------------------------------------- Perform SR 3.2.2.1.

Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours after THERMAL POWER reaching 95% RTP B. Required Action and associated Completion Time not met. B.1 Be in MODE 2. 6 hours 5ACTION e DOC A03 ACTION e.1 ACTION e.2 ACTION e.3 ACTION c.2 DOC M02 3C CINSERT 4 66 3.2.2 Insert Page 3.2.2-2 CTS INSERT 4 CONDITION REQUIRED ACTION COMPLETION TIME B. f1(I) min margin < 0. B.1 Reduce Overtemperature T trip setpoint by TRH multiplied times the f1(I) min margin. 48 hours 4.2.3.2.c.4 6 NHF3.2.2 WOG STS3.2.2-3Rev. 4.0, CTS FH(X,Y)SEQUOYAH UNIT 2Amendment xxx 11SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify NHF is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND [ 31 EFPD thereafter OR In accordance with the Surveillance Frequency Control Program ] 8INSERT 7INSERT 587INSERT 64.2.3.2.c.1 4.2.3.3.a 4.2.3.3.b 899FH min margin > 0 3.2.2 Insert Page 3.2.2-3a CTS INSERT 5 ------------------------------------------------------------NOTE----------------------------------------------------------- Not required to be performed until 12 hours after an equilibrium power level has been achieved, at which a power distribution map can be obtained.

74.2.3.1 3.2.2 Insert Page 3.2.2-3b CTS INSERT 6 -------------------------------NOTE------------------------------If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: FHRM(X,Y) > BHNOM(X,Y) a.Increase HMF(X,Y) by the appropriate factorspecified in the COLR and reverifyFH min margin > 0; orb.Repeat SR 3.2.2.1 prior to the time at which theprojected FH min margin will be < 0.--------------------------------------------------------------------- 4.2.3.2.d 4.2.3.2.d.1 4.2.3.2.d.2 3.2.2 Insert Page 3.2.2-3c CTS INSERT 7 SURVEILLANCE FREQUENCY SR 3.2.2.2 -------------------------------NOTE------------------------------ If two measurements extrapolated to 31 EFPD beyond the most recent measurement yield: FHRM(X,Y) > BHNOM(X,Y) a. Increase HMF(X,Y) by the appropriate factorspecified in the COLR and reverifyf1 (I) min margin > 0; orb.Repeat SR 3.2.2.2 prior to the time at whichthe projected f1 (I) min margin will be < 0.--------------------------------------------------------------------- Verify f1(I) min margin > 0. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND In accordance with the Surveillance Frequency Control Program 84.2.3.2.d 4.2.3.2.c.1 4.2.3.3.a 4.2.3.3.b 4.2.3.2.d.1 4.2.3.2.d.2 JUSTIFICATION FOR DEVIATIONS ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (FH(X,Y)) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS whichreflect the plant specific nomenclature, number, reference, system description,analysis, or licensing basis description.2.ISTS LCO 3.2.2 Required Action A.1.1 states, restore NHFto within limit. ITS3.2.2 will not retain the specific requirement to restore. LCO 3.0.2 Bases states that correction of the entered Condition is an action that may always be considered upon entering ACTIONS. This change is acceptable because the technical requirements have not changed. Restoration of compliance with theLCO is always an available Required Action. The convention in the ITS is to notstate such "restore" options explicitly unless it is the only action or is required for clarity. In this specific application, Required Action A.1.1 is not the only ACTION and a power reduction should be the focus for restoration of FH(X,Y)to within thelimits. Subsequent Required Actions have been renumbered o reflect thisdeletion.3.Required Action A.4 is added to the ITS. CTS 3.2.3 ACTION d requires reductionof the OTT setpoint when FH(X,Y) exceeds the limit in the COLR. SubsequentRequired Actions have been renumbered o reflect this deletion.4.The Completion Times for reducing THERMAL POWER upon discovery thatFH(X,Y) has exceeded its limit are shortened from 4 hours to 2 hours consistentwith the current licensing basis.5.The amount that THERMAL POWER and the Power Range Neutron Flux - HighTrip setpoints are reduced after FH(X,Y) has exceeded its limit are changed toreflect the values in the current licensing basis.6.ITS Conditions A and B have been changed to reflect the CTS ACTIONs for bothFH and/or f1(I) min margins not met.7.ISTS LCO 3.2.2 does not contain a specific provision for changing MODES ifLCO 3.2.2 is not met, other than the generic use of LCO 3.0.4. CTS SR 4.2.3.1states, The provisions of Specification 4.0.4 are not applicable. This allowanceenables SQN to enter the MODE of Applicability with the Surveillance not met or performed. SQN is retaining the allowance to change the MODE of Applicabilitywith the Surveillance not performed by adding a Surveillance Note to retain theallowance.8.ISTS SR 3.2.2.1 and SR 3.2.2.2 have been changed to reflect the CTSevaluation of FH min margin > 0 and f1(I) min margin > 0.9.ISTS SR 3.2.2.1 (and proposed ITS SR 3.2.2.2) provides two options forcontrolling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) HNF B 3.2.2 WOG STS B 3.2.2-1 Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 1 Revision XXX 11B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (HNF ) BASES BACKGROUND The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses. HNF is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, HNF is a measure of the maximum total power produced in a fuel rod. HNF is sensitive to fuel loading patterns, bank insertion, and fuel burnup. HNF typically increases with control bank insertion and typically decreases with fuel burnup. HNF is not directly measurable but is inferred from a power distribution map obtained with the movable incore detector system. Specifically, the results of the three dimensional power distribution map are analyzed by a computer to determine HNF. This factor is calculated at least every 31 EFPD. However, during power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which address directly and continuously measured process variables. The COLR provides peaking factor limits that ensure that the design basis value of the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. The DNB design basis precludes DNB and is met by limiting the minimum local DNB heat flux ratio to [1.3] using the [W3] CHF correlation. All DNB limited transient events are assumed to begin with an HNF value that satisfies the LCO requirements. Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant. FH(X,Y)FH(X,Y) FH(X,Y)FH(X,Y) FH(X,Y)FH(X,Y)1111FH(X,Y)the design limit value using an NRC approved critical heat flux1INSERT 1 2 B 3.2.2 Insert Page B 3.2.2-1a INSERT 1 An FH(X,Y) evaluation requires obtaining an incore flux map in MODE 1. The incore flux map results provide the measured value ()Y,X(FMH of FH(X,Y) for each assembly location (X,Y). The FH ratio (FDHR) is used in order to determine the FH limit for the measured and design power distributions. Then, FHRM(X,Y) = )Y,X(AXIAL/MAP)Y,X(FMMMHwhere MMAP is the maximum allowable peak from the COLR for the measured assembly power distribution at assembly location (X,Y) which accounts for calculational and measurement uncertainties, and )Y,X(AXIALM is the measured ratio of the peak-to-average axial power at assembly location (X,Y). BHDES(X,Y) is a cycle dependent design limit to preserve Departure from Nucleate Boiling(DNB) assumed for initial conditions at the time of limiting transients such as a Loss of Flow Accident (LOFA). BRDES(X,Y) is a cycle dependent design limit to preserve reactor protection system safety limits for DNB requirements. The expression for BHDES(X,Y) is: BHDES(X,Y) = FHRd(X,Y)

MH(X,Y) where: FHRd(X,Y) = )Y,X(AXIAL/MAP)Y,X(FdddH*dMAP is the maximum allowable peak from the COLR for the designassembly power distribution at assembly location (X,Y) whichaccounts for calculational and measurement uncertainties,*)Y,X(AXIALd is the design ratio of the peak-to-average axial powerat assembly location (X,Y),*)Y,X(FdH is the design FH assembly location (X, Y), and *MH(X,Y) is the minimum available margin ratio for initial conditionDNB at the limiting conditions at assembly location (X,Y).2 B 3.2.2 Insert Page B 3.2.2-1b INSERT 1 (continued) The expression for BRDES(X,Y) is: BRDES(X,Y) = FHRd(X,Y)
MHs(X,Y) where: MHs(X,Y) is the minimum available margin ratio for steady state DNB at the limiting conditions at assembly location (X,Y). The reactor core is operating as designed if the measured steady state core power distribution agrees with prediction within statistical variation. This guarantees that the operating limits will preserve the thermal criteria in the applicable safety analyses. The core is operating as designed if the following relationship is satisfied: FHRM(X,Y) BHNOM(X,Y) where: BHNOM(X,Y) is the nominal design radial peaking factor for an assembly at core location (X,Y) increased by an allowance for the expected deviation between the measured and predicted design power distribution. 2 HNF B 3.2.2 WOG STS B 3.2.2-2 Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 1 Revision XXX 11BASES APPLICABLE Limits on HNF preclude core power distributions that exceed the following SAFETY fuel design limits: ANALYSES

a. There must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition,

b. During a large break loss of coolant accident (LOCA), peak cladding temperature (PCT) must not exceed 2200°F, c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm [Ref. 1], and d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn.

For transients that may be DNB limited, the Reactor Coolant System flow and HNF are the core parameters of most importance. The limits on HNF ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the minimum DNBR to the 95/95 DNB criterion of [1.3] using the [W3] CHF correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience a DNB. The allowable HNF limit increases with decreasing power level. This functionality in HNF is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of HNF in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial HNF as a function of power level defined by the COLR limit equation. The LOCA safety analysis indirectly models HNF as an input parameter. The Nuclear Heat Flux Hot Channel Factor (FQ(Z)) and the axial peaking factors are inserted directly into the LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature [Ref. 3]. The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO FH(X,Y)FH(X,Y)FH(X,Y)FH(X,Y) FH(X,Y)1113X,Y, local DNB heat flux ratio to the design limit value using an NRC approved critical heat flux 1limits, FH min margin and f1(I) min margin, )(8) ( 83 HNF B 3.2.2 WOG STS B 3.2.2-3 Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 1 Revision XXX 11BASES

APPLICABLE SAFETY ANALYSES (continued) (QPTR)," LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor )F(HN," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))." HNF and FQ(Z) are measured periodically using the movable incore detector system. Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits. HNF satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO HNF shall be maintained within the limits of the relationship provided in the COLR. The HNF limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB. The limiting value of HNF, described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses. A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of is HNF allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER. APPLICABILITY The HNF limits must be maintained in MODE 1 to preclude core power distributions from exceeding the fuel design limits for DNBR and PCT. Applicability in other modes is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the coolant to require a limit on the distribution of core power. Specifically, the design bases events that are sensitive to HNF in other modes (MODES 2 through 5) have significant margin to DNB, and therefore, there is no need to restrict HNF in these modes. FH(X,Y) FH(X,Y) FH(X,Y)FH(X,Y)FH(X,Y)FH(X,Y)1211X, Y, ZX, Y, ZINSERT 2 indirectly B 3.2.2 Insert Page B 3.2.2-3 INSERT 2 The LCO states that FH(X,Y) shall be less than the limits provided in the COLR. This LCO relationship must be satisfied even if the core is operating at limiting conditions. This requires adjustment to the measured FH(X,Y) to account for limiting conditions and the differences between design and measured conditions. The adjustments are accounted for by comparing FHRM(X,Y) to the limits BHDES(X,Y) and BRDES(X,Y). Therefore, if the FH min margin is >0 and f1(I) min margin >0 the LCO is satisfied.2 HNF B 3.2.2 WOG STS B 3.2.2-4 Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 1 Revision XXX 11BASES

ACTIONS A.1.1 With HNF exceeding its limit, the unit is allowed 4 hours to restore HNF to within its limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring HNF within its power dependent limit. When the HNF limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the HNF value (e.g., static control rod misalignment) are considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs. Thus, the allowed Completion Time of 4 hours provides an acceptable time to restore HNF to within its limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Condition A is modified by a Note that requires that Required Actions A.2 and A.3 must be completed whenever Condition A is entered. Thus, if power is not reduced because this Required Action is completed within the 4 hour time period, Required Action A.2 nevertheless requires another measurement and calculation of HNF within 24 hours in accordance with SR 3.2.2.1. However, if power is reduced below 50% RTP, Required Action A.3 requires that another determination of HNF must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours after reaching or exceeding 95% RTP. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours.

A.1.2.1 and A.1.2.2 If the value of HNF is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to < 50% RTP in accordance with Required Action A.1.2.1 and reduce the Power Range Neutron Flux - High to 55% RTP in accordance with Required Action A.1.2.2. Reducing RTP to < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours for Required Action A.1.2.1 is consistent with those allowed for in Required Action A.1.1 and provides an acceptable time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours for Required Actions A.1.1 and A.1.2.1 are not additive. FH min margin445 5 41FH min marginINSERT 5 INSERT 6 INSERT 7 2 41335allowableINSERT 3 2verified3 INSERT 4 B 3.2.2 Insert Page B 3.2.2-4a INSERT 3 The % FH margin is based on the relationship between FHRM(X,Y) and the limit, BHDES (X,Y), as follows: 100% x Y)BHDES(X,Y)(X,HRF 1 =Margin F %MH If the reactor core is operating as designed, then FHRM(X,Y) is less than BHDES (X,Y) and calculation of %FH margin is not required. If the %FH margin is less than zero, then FHRM(X,Y) is greater than BHDES (X, Y) and the FH(X,Y) limits may not be adequate to prevent exceeding the initial DNB conditions assumed for transients such as a LOFA. BHDES (X,Y) represents the maximum allowable design radial peaking factors which ensures that the initial conditions DNB will be preserved for operation within the LCO limits, and includes allowances for calculational and measurement uncertainties. The FH min margin is the minimum for all core locations examined. INSERT 4 If FH min margin < 0 is restored to within limits prior to completion of the THERMAL POWER reduction in Required Action A.1, compliance of Required Actions A.3 and A.5 must be met.24 B 3.2.2 Insert Page B 3.2.2-4b INSERT 5 from RTP by at least RRH % (where RRH = Thermal power reduction required to compensate for each 1% that FH(X,Y) exceeds its limit) multiplied times the FH min margin INSERT 6 trip setpoints, as specified in TS Table 3.3.1-1 by RRH% multiplied times the FH min margin INSERT 7 by at least RRH% multiplied times the FH min margin 444 HNFB 3.2.2 WOG STSB3.2.2-5Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 1 Revision XXX 11BASES ACTIONS (continued) The allowed Completion Time of 72 hours to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System. A.2 Once the power level has been reduced to < 50% RTP per Required Action A.1.2.1, an incore flux map (SR 3.2.2.1) must be obtained and the measured value of HNF verified not to exceed the allowed limit at the lower power level. The unit is provided 20 additional hours to perform this task over and above the 4 hours allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours is acceptable because of the increase in the DNB margin, which is obtained at lower power levels, and the low probability of having a DNB limiting event within this 24 hour period. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain the incore flux map, perform the required calculations, and evaluate HNF. A.3 Verification that HNF is within its specified limits after an out of limit occurrence ensures that the cause that led to the HNF exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the HNF limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is 95% RTP. This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action. B.1 When Required Actions A.1.1 through A.3 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems. FH min margin is verified > 0FHmin marginFHmin margin is > 0FH min marginFH min margin1114INSERT 8 4322 2 INSERT 945 45, and B.1, 4C 2allowable3 >0INSERT 10 B 3.2.2 Insert Page B 3.2.2-5a INSERT 8 by at least RRH% multiplied times the FH min margin INSERT 9 A.4 If the value of FHRM(X,Y) is not restored to within its specified limit, Overtemperarture T K1 (OTT K1) term is required to be reduced by at least TRH multiplied times the FH min margin. The value of TRH is provided in the COLR. Completing Required Action A.4 ensures protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during an anticipated operational occurrence. Also, completing Required Action A.4 within the allowed Completion Time of 48 hours is sufficient considering the small likelihood of a limiting transient in this time period. 44 B 3.2.2 Insert Page B 3.2.2-5b INSERT 10 B.1 The %f1(I) margin is based on the relationship between FHRM(X,Y) and the limit, BRDES (X,Y), as follows:

% f(I)Margin = 1 FHR(X,Y)BRDES(X,Y) x 100%1M If the reactor core is operating as designed, then FHRM(X,Y) is less than BRDES (X,Y) and calculation of %f1(I) margin is not required. If the %f1(I) margin is less than zero, then FHRM(X,Y) is greater than BRDES (X, Y) and the OTT setpoint limits may not be adequate to prevent exceeding DNB requirements. BRDES (X,Y) represents the maximum allowable design radial peaking factors which ensure that the steady state DNBR limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties Required Action B.1 requires the reduction of the OTT K1 term by at least TRH multiplied by the f1(I) min margin. TRH is the amount of OTT K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR. Completing Required Action B.1 within the allowed Completion Time of 48 hours, restricts FH(X,Y) such that even if a transient occurred, DNB requirements are met. The f1(I) min margin is the minimum % of f1(I) margin for all core locations examined. 2 HNFB 3.2.2 WOG STSB3.2.2-6Rev. 4.0, FH(X,Y)1SEQUOYAH UNIT 1 Revision XXX 1BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of HNF is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of HNF from the measured flux distributions. The measured value of HNF must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the HNF limit. After each refueling, HNF must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that HNF limits are met at the beginning of each fuel cycle. [ The 31 EFPD Frequency is acceptable because the power distribution changes relatively slowly over this amount of fuel burnup. Accordingly, this Frequency is short enough that the HNF limit cannot be exceeded for any significant period of operation. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1.Regulatory Guide 1.77, Rev. [0], May 1974.2.10 CFR 50, Appendix A, GDC 26.

3.10 CFR 50.46.1INSERT 12 4467INSERT 11 B 3.2.2 Insert Page B 3.2.2-6a INSERT 11 SR 3.2.2.1 and SR 3.2.2.2 are modified by a Note. It states that, "Not required to be performed until 12 hours after an equilibrium power level has been achieved at which a power distribution map can be obtained." SR 3.2.2.1 and SR 3.2.2.2 require using the incore detector system to provide the necessary data to create a power distribution map. To provide the necessary data, MODE 1 needs to be entered, power escalated, stabilized and equilibrium conditions established at some higher power level. These surveillances could not be satisfactorily performed if the requirement for performance of the Surveillances was included in MODE 2 prior to entering MODE 1. In a reload core,HMF(X,Y)could not have previously been measured, therefore, there is a Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of HMF(X,Y) is made at a lower power level at which adequate margin is available before going to 100% RTP. 4 B 3.2.2 Insert Page B 3.2.2-6b INSERT 12 SR 3.2.2.1 and SR 3.2.2.2 In addition to ensuring via Surveillance that the nuclear enthalpy rise hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in )Y,X(FMH for the last two measurements out to 31 EFPD beyond the most recent measurement. If the extrapolation yields an FHRM(X,Y) > BHNOM(X,Y), further consideration is required. The implications of these extrapolations are considered separately for BHDES(X,Y) and BRDES(X,Y) limits. If the extrapolations of )Y,X(FMH are unfavorable, additional actions must be taken. These actions are to meet the FH(X,Y) limit with the last HMF(X,Y) increased by the appropriate factor specified in the COLR or to evaluateHMF(X,Y) prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data. Extrapolation is not required for the initial flux map taken after reaching equilibrium conditions following a refueling outage since the initial flux map establishes the baseline measurement for future trending. 4 HNFB 3.2.2 WOG STSB3.2.2-1Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 2 Revision XXX 11B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (HNF ) BASES BACKGROUND The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses. HNF is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, HNF is a measure of the maximum total power produced in a fuel rod. HNF is sensitive to fuel loading patterns, bank insertion, and fuel burnup. HNF typically increases with control bank insertion and typically decreases with fuel burnup. HNF is not directly measurable but is inferred from a power distribution map obtained with the movable incore detector system. Specifically, the results of the three dimensional power distribution map are analyzed by a computer to determine HNF. This factor is calculated at least every 31 EFPD. However, during power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which address directly and continuously measured process variables. The COLR provides peaking factor limits that ensure that the design basis value of the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. The DNB design basis precludes DNB and is met by limiting the minimum local DNB heat flux ratio to [1.3] using the [W3] CHF correlation. All DNB limited transient events are assumed to begin with an HNF value that satisfies the LCO requirements. Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant. FH(X,Y)FH(X,Y) FH(X,Y)FH(X,Y) FH(X,Y)FH(X,Y)1111FH(X,Y)the design limit value using an NRC approved critical heat flux1INSERT 1 2 B 3.2.2 Insert Page B 3.2.2-1a INSERT 1 An FH(X,Y) evaluation requires obtaining an incore flux map in MODE 1. The incore flux map results provide the measured value ()Y,X(FMH of FH(X,Y) for each assembly location (X,Y). The FH ratio (FDHR) is used in order to determine the FH limit for the measured and design power distributions. Then, FHRM(X,Y) = )Y,X(AXIAL/MAP)Y,X(FMMMHwhere MMAP is the maximum allowable peak from the COLR for the measured assembly power distribution at assembly location (X,Y) which accounts for calculational and measurement uncertainties, and )Y,X(AXIALM is the measured ratio of the peak-to-average axial power at assembly location (X,Y). BHDES(X,Y) is a cycle dependent design limit to preserve Departure from Nucleate Boiling(DNB) assumed for initial conditions at the time of limiting transients such as a Loss of Flow Accident (LOFA). BRDES(X,Y) is a cycle dependent design limit to preserve reactor protection system safety limits for DNB requirements. The expression for BHDES(X,Y) is: BHDES(X,Y) = FHRd(X,Y)

MH(X,Y) where: FHRd(X,Y) = )Y,X(AXIAL/MAP)Y,X(FdddH*dMAP is the maximum allowable peak from the COLR for the designassembly power distribution at assembly location (X,Y) whichaccounts for calculational and measurement uncertainties,*)Y,X(AXIALd is the design ratio of the peak-to-average axial powerat assembly location (X,Y),*)Y,X(FdH is the design FH assembly location (X, Y), and *MH(X,Y) is the minimum available margin ratio for initial conditionDNB at the limiting conditions at assembly location (X,Y).2 B 3.2.2 Insert Page B 3.2.2-1b INSERT 1 (continued) The expression for BRDES(X,Y) is: BRDES(X,Y) = FHRd(X,Y)
MHs(X,Y) where: MHs(X,Y) is the minimum available margin ratio for steady state DNB at the limiting conditions at assembly location (X,Y). The reactor core is operating as designed if the measured steady state core power distribution agrees with prediction within statistical variation. This guarantees that the operating limits will preserve the thermal criteria in the applicable safety analyses. The core is operating as designed if the following relationship is satisfied: FHRM(X,Y) BHNOM(X,Y) where: BHNOM(X,Y) is the nominal design radial peaking factor for an assembly at core location (X,Y) increased by an allowance for the expected deviation between the measured and predicted design power distribution. 2 HNFB 3.2.2 WOG STSB3.2.2-2Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 2 Revision XXX 11BASES APPLICABLE Limits on HNF preclude core power distributions that exceed the following SAFETY fuel design limits: ANALYSESa.There must be at least 95% probability at the 95% confidence level(the 95/95 DNB criterion) that the hottest fuel rod in the core does notexperience a DNB condition,b.During a large break loss of coolant accident (LOCA), peak claddingtemperature (PCT) must not exceed 2200°F,c.During an ejected rod accident, the energy deposition to the fuelmust not exceed 280 cal/gm [Ref. 1], andd.Fuel design limits required by GDC 26 (Ref. 2) for the condition whencontrol rods must be capable of shutting down the reactor with aminimum required SDM with the highest worth control rod stuck fullywithdrawn.For transients that may be DNB limited, the Reactor Coolant System flow and HNF are the core parameters of most importance. The limits on HNF ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the minimum DNBR to the 95/95 DNB criterion of [1.3] using the [W3] CHF correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience a DNB. The allowable HNF limit increases with decreasing power level. This functionality in HNF is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of HNF in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial HNF as a function of power level defined by the COLR limit equation. The LOCA safety analysis indirectly models HNF as an input parameter. The Nuclear Heat Flux Hot Channel Factor (FQ(Z)) and the axial peaking factors are inserted directly into the LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature [Ref. 3]. The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO FH(X,Y)FH(X,Y)FH(X,Y)FH(X,Y) FH(X,Y)1113X,Y, local DNB heat flux ratio to the design limit value using an NRC approved critical heat flux 1limits, FH min margin and f1(I) min margin, )(8) ( 83 HNFB 3.2.2 WOG STSB3.2.2-3Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 2 Revision XXX 11BASES APPLICABLE SAFETY ANALYSES (continued) (QPTR)," LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor )F(HN," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))." HNF and FQ(Z) are measured periodically using the movable incore detector system. Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits. HNF satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO HNF shall be maintained within the limits of the relationship provided in the COLR. The HNF limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB. The limiting value of HNF, described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses. A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of is HNF allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER. APPLICABILITY The HNF limits must be maintained in MODE 1 to preclude core power distributions from exceeding the fuel design limits for DNBR and PCT.

Applicability in other modes is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the coolant to require a limit on the distribution of core power. Specifically, the design bases events that are sensitive to HNF in other modes (MODES 2 through 5) have significant margin to DNB, and therefore, there is no need to restrict HNF in these modes. FH(X,Y) FH(X,Y) FH(X,Y)FH(X,Y)FH(X,Y)FH(X,Y)1211X, Y, ZX, Y, ZINSERT 2 indirectly B 3.2.2 Insert Page B 3.2.2-3 INSERT 2 The LCO states that FH(X,Y) shall be less than the limits provided in the COLR. This LCO relationship must be satisfied even if the core is operating at limiting conditions. This requires adjustment to the measured FH(X,Y) to account for limiting conditions and the differences between design and measured conditions. The adjustments are accounted for by comparing FHRM(X,Y) to the limits BHDES(X,Y) and BRDES(X,Y). Therefore, if the FH min margin is >0 and f1(I) min margin >0 the LCO is satisfied.2 HNF B 3.2.2 WOG STS B 3.2.2-4 Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 2 Revision XXX 11BASES

ACTIONS A.1.1 With HNF exceeding its limit, the unit is allowed 4 hours to restore HNF to within its limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring HNF within its power dependent limit. When the HNF limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the HNF value (e.g., static control rod misalignment) are considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs. Thus, the allowed Completion Time of 4 hours provides an acceptable time to restore HNF to within its limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Condition A is modified by a Note that requires that Required Actions A.2 and A.3 must be completed whenever Condition A is entered. Thus, if power is not reduced because this Required Action is completed within the 4 hour time period, Required Action A.2 nevertheless requires another measurement and calculation of HNF within 24 hours in accordance with SR 3.2.2.1. However, if power is reduced below 50% RTP, Required Action A.3 requires that another determination of HNF must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours after reaching or exceeding 95% RTP. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours.

A.1.2.1 and A.1.2.2 If the value of HNF is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to < 50% RTP in accordance with Required Action A.1.2.1 and reduce the Power Range Neutron Flux - High to 55% RTP in accordance with Required Action A.1.2.2. Reducing RTP to < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours for Required Action A.1.2.1 is consistent with those allowed for in Required Action A.1.1 and provides an acceptable time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours for Required Actions A.1.1 and A.1.2.1 are not additive. FH min margin445 5 41FH min marginINSERT 5 INSERT 6 INSERT 7 2 41335allowableINSERT 3 2verified3 INSERT 4 B 3.2.2 Insert Page B 3.2.2-4a INSERT 3 The % FH margin is based on the relationship between FHRM(X,Y) and the limit, BHDES (X,Y), as follows: 100% x Y)BHDES(X,Y)(X,HRF 1 =Margin F %MH If the reactor core is operating as designed, then FHRM(X,Y) is less than BHDES (X,Y) and calculation of %FH margin is not required. If the %FH margin is less than zero, then FHRM(X,Y) is greater than BHDES (X, Y) and the FH(X,Y) limits may not be adequate to prevent exceeding the initial DNB conditions assumed for transients such as a LOFA. BHDES (X,Y) represents the maximum allowable design radial peaking factors which ensures that the initial conditions DNB will be preserved for operation within the LCO limits, and includes allowances for calculational and measurement uncertainties. The FH min margin is the minimum for all core locations examined. INSERT 4 If FH min margin < 0 is restored to within limits prior to completion of the THERMAL POWER reduction in Required Action A.1, compliance of Required Actions A.3 and A.5 must be met.24 B 3.2.2 Insert Page B 3.2.2-4b INSERT 5 from RTP by at least RRH % (where RRH = Thermal power reduction required to compensate for each 1% that FH(X,Y) exceeds its limit) multiplied times the FH min margin INSERT 6 trip setpoints, as specified in TS Table 3.3.1-1 by RRH% multiplied times the FH min margin INSERT 7 by at least RRH% multiplied times the FH min margin 444 HNFB 3.2.2 WOG STSB3.2.2-5Rev. 4.0, FH(X,Y)SEQUOYAH UNIT 2 Revision XXX 11BASES ACTIONS (continued) The allowed Completion Time of 72 hours to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System. A.2 Once the power level has been reduced to < 50% RTP per Required Action A.1.2.1, an incore flux map (SR 3.2.2.1) must be obtained and the measured value of HNF verified not to exceed the allowed limit at the lower power level. The unit is provided 20 additional hours to perform this task over and above the 4 hours allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours is acceptable because of the increase in the DNB margin, which is obtained at lower power levels, and the low probability of having a DNB limiting event within this 24 hour period. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain the incore flux map, perform the required calculations, and evaluate HNF. A.3 Verification that HNF is within its specified limits after an out of limit occurrence ensures that the cause that led to the HNF exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the HNF limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is 95% RTP. This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action. B.1 When Required Actions A.1.1 through A.3 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems. FH min margin is verified > 0FHmin marginFHmin margin is > 0FH min marginFH min margin1114INSERT 8 4322 2 INSERT 945 45, and B.1, 4C 2allowable3 >0INSERT 10 B 3.2.2 Insert Page B 3.2.2-5a INSERT 8 by at least RRH% multiplied times the FH min margin INSERT 9 A.4 If the value of FHRM(X,Y) is not restored to within its specified limit, Overtemperarture T K1 (OTT K1) term is required to be reduced by at least TRH multiplied times the FH min margin. The value of TRH is provided in the COLR. Completing Required Action A.4 ensures protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during an anticipated operational occurrence. Also, completing Required Action A.4 within the allowed Completion Time of 48 hours is sufficient considering the small likelihood of a limiting transient in this time period. 44 B 3.2.2 Insert Page B 3.2.2-5b INSERT 10 B.1 The %f1(I) margin is based on the relationship between FHRM(X,Y) and the limit, BRDES (X,Y), as follows:

% f(I)Margin = 1 FHR(X,Y)BRDES(X,Y) x 100%1M If the reactor core is operating as designed, then FHRM(X,Y) is less than BRDES (X,Y) and calculation of %f1(I) margin is not required. If the %f1(I) margin is less than zero, then FHRM(X,Y) is greater than BRDES (X, Y) and the OTT setpoint limits may not be adequate to prevent exceeding DNB requirements. BRDES (X,Y) represents the maximum allowable design radial peaking factors which ensure that the steady state DNBR limit will be preserved for operation within the LCO limits, including allowances for calculational and measurement uncertainties Required Action B.1 requires the reduction of the OTT K1 term by at least TRH multiplied by the f1(I) min margin. TRH is the amount of OTT K1 setpoint reduction required to compensate for each 1% that FH(X,Y) exceeds the limit provided in the COLR. Completing Required Action B.1 within the allowed Completion Time of 48 hours, restricts FH(X,Y) such that even if a transient occurred, DNB requirements are met. The f1(I) min margin is the minimum % of f1(I) margin for all core locations examined. 2 HNFB 3.2.2 WOG STSB3.2.2-6Rev. 4.0, FH(X,Y)1SEQUOYAH UNIT 2Revision XXX 1BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of HNF is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of HNF from the measured flux distributions. The measured value of HNF must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the HNF limit. After each refueling, HNF must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that HNF limits are met at the beginning of each fuel cycle. [ The 31 EFPD Frequency is acceptable because the power distribution changes relatively slowly over this amount of fuel burnup. Accordingly, this Frequency is short enough that the HNF limit cannot be exceeded for any significant period of operation. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1.Regulatory Guide 1.77, Rev. [0], May 1974.2.10 CFR 50, Appendix A, GDC 26.

3.10 CFR 50.46.1INSERT 12 4467INSERT 11 B 3.2.2 Insert Page B 3.2.2-6a INSERT 11 SR 3.2.2.1 and SR 3.2.2.2 are modified by a Note. It states that, "Not required to be performed until 12 hours after an equilibrium power level has been achieved at which a power distribution map can be obtained." SR 3.2.2.1 and SR 3.2.2.2 require using the incore detector system to provide the necessary data to create a power distribution map. To provide the necessary data, MODE 1 needs to be entered, power escalated, stabilized and equilibrium conditions established at some higher power level. These surveillances could not be satisfactorily performed if the requirement for performance of the Surveillances was included in MODE 2 prior to entering MODE 1. In a reload core,HMF(X,Y)could not have previously been measured, therefore, there is a Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of HMF(X,Y) is made at a lower power level at which adequate margin is available before going to 100% RTP. 4 B 3.2.2 Insert Page B 3.2.2-6b INSERT 12 SR 3.2.2.1 and SR 3.2.2.2 In addition to ensuring via Surveillance that the nuclear enthalpy rise hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in )Y,X(FMH for the last two measurements out to 31 EFPD beyond the most recent measurement. If the extrapolation yields an FHRM(X,Y) > BHNOM(X,Y), further consideration is required. The implications of these extrapolations are considered separately for BHDES(X,Y) and BRDES(X,Y) limits. If the extrapolations of )Y,X(FMH are unfavorable, additional actions must be taken. These actions are to meet the FH(X,Y) limit with the last HMF(X,Y) increased by the appropriate factor specified in the COLR or to evaluateHMF(X,Y) prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data. Extrapolation is not required for the initial flux map taken after reaching equilibrium conditions following a refueling outage since the initial flux map establishes the baseline measurement for future trending. 4 JUSTIFICATION FOR DEVIATIONS ITS 3.2.2, BASES, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (FH(X,Y)) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS 3.2.2 LCO and Action A Bases have been modified to add details associated with the relationship between FHRM(X,Y) and BHDES(X,Y) in accordance with NRC Safety Evaluation dated April 27, 1997 (ML013320456).

3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is changed to reflect the current licensing basis. 4. Changes have been made to be consistent with changes made to the Specification.

5. The ISTS 3.2.2 Bases for A.1.1, 2nd paragraph, contains in part, "Required Action A.2 nevertheless requires another measurement and calculation of HNF within 24 hours in accordance with SR 3.2.2.1." The last paragraph contains a similar statement, " In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours." SQN is deleting the redundant statement in last paragraph. 6. ISTS SR 3.2.2.1 provides two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies for ITS SR 3.2.2.1 under the Surveillance Frequency Control Program. 7. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

8. Editorial change made for clarification.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.2.2, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (FH(X,Y)) Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 3 ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.2.3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER* ACTION: a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the COLR; 1. Either restore the indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours. b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

    *See Special Test Exception 3.10.2    October 4, 1995 SEQUOYAH - UNIT 1 3/4 2-1 Amendment No. 19, 155, 213 Page 1 of 6 LCO 3.2.3 Applicability L01L02A04A02M01 in % flux difference unitsA03A03ACTION A A02A02 A01ITS ITS 3.2.3 POWER DISTRIBUTION LIMITS   SURVEILLANCE REQUIREMENTS   4.2.1.1  The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:    a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE.

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits.

December 2, 1986 SEQUOYAH - UNIT 1 3/4 2-2 Amendment No. 51 SR 3.2.3.1 LCO 3.2.3 Note L03Page 2 of 6 LA01In accordance with the Surveillance Frequency Control Program A02A02A02 A01ITS ITS 3.2.3 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

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December 23, 1982 SEQUOYAH - UNIT 1 3/4 2-3 Amendment No. 19 Page 3 of 6 A01ITS ITS 3.2.3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER*. ACTION: a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the COLR; 1. Either restore the indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours. b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

     *See Special Test Exception 3.10.2

October 4, 1995 SEQUOYAH - UNIT 2 3/4 2-1 Amendment Nos. 21, 146, 203 Page 4 of 6 LCO 3.2.3 Applicability ACTION A L01L02A04M01 in % flux difference unitsA03A03A02A02 A01ITS ITS 3.2.3 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits.

December 2, 1986 SEQUOYAH - UNIT 2 3/4 2-2 Amendment No. 43 Page 5 of 6 SR 3.2.3.1 LCO 3.2.3 Note L03LA01In accordance with the Surveillance Frequency Control Program A02A02A02 A01ITS ITS 3.2.3 This page intentionally deleted March 30, 1992 SEQUOYAH - UNIT 2 3/4 2-3 Amendment Nos. 21, 146 Page 6 of 6 DISCUSSION OF CHANGES ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.2.1 states "The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the COLR." CTS 3.2.1 ACTION a provides ACTIONs to take when the indicated AFD is outside the limits. CTS 4.2.1.1 requires a determination that the indicated AFD is within limits. CTS 4.2.1.2 states that the indicated AFD shall be considered outside the limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits. ITS LCO 3.2.3 states "The AFD in % flux difference units shall be maintained within the limits specified in the COLR." ITS LCO 3.2.3 is modified by a Note specifying when AFD is considered to be outside the limits. ITS SR 3.2.3.1 requires verification that AFD is within limits. This changes the CTS by deleting "indicated and adding "% flux difference units" to the LCO statement. The purpose of CTS 3.2.1 is to ensure the AFD remains within the limits specified in the COLR. AFD is the difference in normalized flux signals between the top and bottom excore detectors, therefore, this is a presentation change. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 3.2.1 Applicability contains a footnote (footnote *) which states "See Special Test Exception 3.10.2." ITS 3.2.3 Applicability does not contain this footnote. This changes the CTS by not including Footnote*. The purpose of Footnote

is to alert the Technical Specification user that a Special Test Exception exists that may modify the Applicability of this Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative because it does not result in a technical change to the CTS. A04 CTS 3.2.1 ACTION b states "THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR." ITS 3.2.3 does not contain a similar requirement. This changes the CTS by eliminating a prohibition contained in the CTS. This change is acceptable because the requirements have not changed. CTS 3.0.4 and ITS 3.0.4 prohibit entering the MODE of Applicability of a Technical Specification unless the requirements of the LCO are met. CTS 3.2.1 and ITS 3.2.3 are applicable in MODE 1 with THERMAL POWER > 50% RTP (CTS) and 50 RTP (ITS). Therefore, both the CTS and ITS prohibit exceeding DISCUSSION OF CHANGES ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 2 of 4 50% RTP without the LCO requirements being met. CTS 3.2.1 ACTION b is duplicative of CTS 3.0.4 and ITS 3.0.4 and its elimination does not make a technical change to the Specification. This change is designated as an administrative change because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 3.2.1 is applicable in MODE 1 with THERMAL POWER > 50% RTP. ITS 3.2.3 is applicable in MODE 1 with THERMAL POWER 50% RTP. This changes the CTS by requiring LCO 3.2.3 to be met when THERMAL POWER is equal to 50 % RTP. The purpose of CTS 3.2.1 is to maintain the AFD within the limits specified in the COLR. When AFD is not within limits, CTS 3.2.1 ACTION a.2, requires reducing THERMAL POWER to less than 50% RTP. This change is acceptable because it aligns the Applicability to the Required Actions. The CTS and ITS Required Action is to reduce THERMAL POWER to less than 50% RTP. When the THERMAL POWER is reduced to this value, it places the core in a condition outside of the Applicability of the LCO. Therefore, changing the Applicability from in MODE 1 with THERMAL POWER > 50% RTP to MODE 1 with THERMAL POWER 50% RTP has no affect on the LCO. This change is designated as more restrictive because it provides additional requirements to the Applicability. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.2.1.1.a requires monitoring the indicated AFD for each OPERABLE excore channel at least once per 7days. ITS SR 3.2.3.1 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequency for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance DISCUSSION OF CHANGES ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 3 of 4 Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.2.1 ACTION a.1 requires with the AXIAL FLUX DIFFERENCE (AFD) outside of the limits, to restore the indicated AFD to within the limits within 15 minutes. ITS 3.2.3 does not include a Required Action to restore the indicated AFD to within the limits within 15 minutes. This changes the CTS by not including a specific requirement to restore the AFD to within limits. The purpose of CTS 3.2.1 is to maintain the AFD within the limits specified in the COLR. This change is acceptable because the requirement to restore the AFD to within limits has not changed. ITS 3.2.3 allows a Completion Time of 30 minutes to reduce THERMAL POWER to < 50% RTP. During the time that power is being reduced, AFD can be restored to within limits. Per ITS LCO 3.0.2, if the LCO is met prior to expiration of the Completion Time, completion of the Required Actions is not required. This allowance also is provided in CTS 3.0.2. Therefore, restoration of AFD is always an option and a specific ACTION is not required. This change is designated as less restrictive because additional Completion Time is provided that was not provided in the CTS. L02 (Category 4 - Relaxation of Required Action) CTS 3.2.1 ACTION a.2 states that with the indicated AFD outside of the limits specified in the COLR, reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours. ITS 3.2.3 ACTION A only requires THERMAL POWER to be reduced to less than 50% RTP. This changes the CTS by eliminating the requirement to reduce the Power Range Neutron Flux - High trip setpoints to 55 % of RTP within the next 4 hours. The purpose of CTS 3.2.1 ACTION a.2 is to reduce THERMAL POWER to the point at which the LCO is met if AFD is not restored within its limit. With the AFD meeting the Technical Specification requirements, further actions are not required to ensure that the assumptions of the safety analyses are met. Increases in THERMAL POWER are governed by ITS LCO 3.0.4, which requires the LCO to be met prior to entering a MODE or other specified condition in which the LCO applies. Therefore, power increases are prohibited while avoiding the risk of changing Reactor Trip System setpoints during operation. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L03 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.2.1.1.a requires the monitoring of the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE. CTS 4.2.1.1.b DISCUSSION OF CHANGES ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 4 of 4 requires the monitoring and logging of the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging. This changes the CTS by eliminating all AFD Surveillance Frequencies based on the OPERABILITY of the AFD Monitor Alarm. The purpose of ITS 3.2.3 is to ensure that AFD is within its limit. This change is acceptable because the remaining Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of monitoring AFD when the AFD Monitor Alarm is inoperable is unnecessary as inoperability of the alarm does not increase the probability that AFD is outside of its limit. The AFD Monitor Alarm is for indication only. Its use is not credited in any safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) AFD (RAOC Methodology) 3.2.3B Westinghouse STS 3.2.3B-1 Rev. 4.0 Amendment XXX SEQUOYAH UNIT 1 11CTS 23.2 POWER DISTRIBUTION LIMITS 3.2.3B AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) LCO 3.2.3B The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

--------------------------------------------NOTE---------------------------------------------  The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.  --------------------------------------------------------------------------------------------------

APPLICABILITY: MODE 1 with THERMAL POWER 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER to < 50% RTP. 30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] 1333.2.1 Applicability ACTION A.2 4.2.1.2 4.2.1.1.a 1 AFD (RAOC Methodology) 3.2.3B Westinghouse STS 3.2.3B-1 Rev. 4.0 Amendment XXX SEQUOYAH UNIT 2 11CTS 23.2 POWER DISTRIBUTION LIMITS 3.2.3B AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) LCO 3.2.3B The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

--------------------------------------------NOTE---------------------------------------------  The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.  --------------------------------------------------------------------------------------------------

APPLICABILITY: MODE 1 with THERMAL POWER 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER to < 50% RTP. 30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] 1333.2.1 Applicability ACTION A.2 4.2.1.2 4.2.1.1.a 1 JUSTIFICATION FOR DEVIATIONS ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The type of Methodology (Relaxed Axial Offset Control (RAOC)) and the Specification designator "B" are deleted since they are unnecessary (only one AFD Specification is used in the Sequoyah Nuclear (SQN) Plant ITS.) This information is provided in NUREG-1431, Rev. 4.0, to assist in indentifying the appropriate Specification to be used as a model for the plant specific ITS conversion, but serves no purpose in a plant specific implementation. In addition, the Constant Axial Offset Control (CAOC) methodology Specification (ISTS 3.2.3A) is not used and is not shown.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS SR 3.2.3.1 provides two options for controlling the Frequency of the Surveillance Requirement. SQN is proposing to control the Surveillance Frequency under the Surveillance Frequency Control Program.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-1 Rev. 4.0 211SEQUOYAH UNIT 1 Revision XXX B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3B AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC Methodology) BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control. RAOC is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity. The AFD is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels is outside its specified limits. Although the RAOC defines limits that must be met to satisfy safety analyses, typically an operating scheme, Constant Axial Offset Control (CAOC), is used to control axial power distribution in day to day operation (Ref. 1). CAOC requires that the AFD be controlled within a narrow tolerance band around a burnup dependent target to minimize the variation of axial peaking factors and axial xenon distribution during unit maneuvers. The CAOC operating space is typically smaller and lies within the RAOC operating space. Control within the CAOC operating space constrains the variation of axial xenon distributions and axial power distributions. RAOC calculations assume a wide range of xenon distributions and then confirm that the resulting power distributions satisfy the requirements of the accident analyses. 1111 AFD (RAOC Methodology) B 3.2.3B Westinghouse STSB 3.2.3B-2 Rev. 4.0 211SEQUOYAH UNIT 1 Revision XXX BASES APPLICABLE The AFD is a measure of the axial power distribution skewing to either the SAFETY top or bottom half of the core. The AFD is sensitive to many core related ANALYSES parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration. The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements. The RAOC methodology (Ref. 2) establishes a xenon distribution library with tentatively wide AFD limits. One dimensional axial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements. The limits on the AFD ensure that the Heat Flux Hot Channel Factor (FQ(Z)) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition 2, 3, or 4 events. This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition 4 event is the LOCA. The most important Condition 3 event is the loss of flow accident. The most important Condition 2 events are uncontrolled bank withdrawal and boration or dilution accidents. Condition 2 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower T and Overtemperature T trip setpoints. The limits on the AFD satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes. Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as % flux or %I. 1 and 212X, Y, 22s 6, 6Enclosure 2, Volume 7, Rev. , Page 191 of 249Enclosure 2, Volume 7, Rev. , Page 191 of 249A Condition 4 event significantly affected by the initial axial power distribution, as indicated by AFD, is the LOCA. A Condition 3 event significantly affected by AFD is the Complete Loss of RCS Flow event. A Condition 2 event significantly affected by AFD is the Uncontrolled RCCA Bank Withdrawal at Power event. (Ref.2)1and3(Ref.1) AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-3 Rev. 4.0 211SEQUOYAH UNIT 1 Revision XXX BASES LCO (continued) The AFD limits are provided in the COLR. Figure B 3.2.3B-1 shows typical RAOC AFD limits. The AFD limits for RAOC do not depend on the target flux difference. However, the target flux difference may be used to minimize changes in the axial power distribution. Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits. APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis. For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES. ACTIONSA.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to < 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems. SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits. [ The Surveillance Frequency of 7 days is adequate considering that the AFD is monitored by a computer and any deviation from requirements is alarmed. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] INSERT 1151343 B 3.2.3 Insert Page B 3.2.3-3 INSERT 1 The AFD limits resulting from analysis of core power distributions relative to the initial condition peaking limits comprise a power-dependant envelope of acceptable AFD values. During steady-state operation, the core normally is controlled to a target AFD within a narrow (approximately +/- 5% AFD) band. However, the limiting AFD values may be somewhat greater than the extremes of the normal operating band. 1 AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-4 Rev. 4.0 1Revision XXX1SEQUOYAH UNIT 12BASES REFERENCES 1. WCAP-8403 (nonproprietary), "Power Distribution Control and LoadFollowing Procedures," Westinghouse Electric Corporation,September 1974.2.R. W. Miller et al., "Relaxation of Constant Axial Offset Control: FQSurveillance Technical Specification," WCAP-10217(NP), June 1983.3.FSAR, Chapter [15].UFSAR, Section 4.3.2.BAW10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. 222Enclosure 2, Volume 7, Rev. , Page 194 of 249Enclosure 2, Volume 7, Rev. , Page 194 of 24932.UFSAR,Chapter15.1 AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-5 Rev. 4.0 1Revision XXX1SEQUOYAH UNIT 125 AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-1 Rev. 4.0 211SEQUOYAH UNIT 2 Revision XXX B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3B AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC Methodology) BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control. RAOC is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity. The AFD is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels is outside its specified limits. Although the RAOC defines limits that must be met to satisfy safety analyses, typically an operating scheme, Constant Axial Offset Control (CAOC), is used to control axial power distribution in day to day operation (Ref. 1). CAOC requires that the AFD be controlled within a narrow tolerance band around a burnup dependent target to minimize the variation of axial peaking factors and axial xenon distribution during unit maneuvers. The CAOC operating space is typically smaller and lies within the RAOC operating space. Control within the CAOC operating space constrains the variation of axial xenon distributions and axial power distributions. RAOC calculations assume a wide range of xenon distributions and then confirm that the resulting power distributions satisfy the requirements of the accident analyses. 1111 AFD (RAOC Methodology) B 3.2.3B Westinghouse STSB 3.2.3B-2 Rev. 4.0 211SEQUOYAH UNIT 2 Revision XXX BASES APPLICABLE The AFD is a measure of the axial power distribution skewing to either the SAFETY top or bottom half of the core. The AFD is sensitive to many core related ANALYSES parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration. The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements. The RAOC methodology (Ref. 2) establishes a xenon distribution library with tentatively wide AFD limits. One dimensional axial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements. The limits on the AFD ensure that the Heat Flux Hot Channel Factor (FQ(Z)) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition 2, 3, or 4 events. This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition 4 event is the LOCA. The most important Condition 3 event is the loss of flow accident. The most important Condition 2 events are uncontrolled bank withdrawal and boration or dilution accidents. Condition 2 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower T and Overtemperature T trip setpoints. The limits on the AFD satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes. Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as % flux or %I. 1 and 212X, Y, 22s 6, 6Enclosure 2, Volume 7, Rev. , Page 197 of 249Enclosure 2, Volume 7, Rev. , Page 197 of 249A Condition 4 event significantly affected by the initial axial power distribution, as indicated by AFD, is the LOCA. A Condition 3 event significantly affected by AFD is the Complete Loss of RCS Flow event. A Condition 2 event significantly affected by AFD is the Uncontrolled RCCA Bank Withdrawal at Power event. (Ref.2)1and3(Ref.1) AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-3 Rev. 4.0 211SEQUOYAH UNIT 2 Revision XXX BASES LCO (continued) The AFD limits are provided in the COLR. Figure B 3.2.3B-1 shows typical RAOC AFD limits. The AFD limits for RAOC do not depend on the target flux difference. However, the target flux difference may be used to minimize changes in the axial power distribution. Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits. APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis. For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES. ACTIONSA.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to < 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems. SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits. [ The Surveillance Frequency of 7 days is adequate considering that the AFD is monitored by a computer and any deviation from requirements is alarmed. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] INSERT 1151343 B 3.2.3 Insert Page B 3.2.3-3 INSERT 1 The AFD limits resulting from analysis of core power distributions relative to the initial condition peaking limits comprise a power-dependant envelope of acceptable AFD values. During steady-state operation, the core normally is controlled to a target AFD within a narrow (approximately +/- 5% AFD) band. However, the limiting AFD values may be somewhat greater than the extremes of the normal operating band. 1 AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-4 Rev. 4.0 1Revision XXX1SEQUOYAH UNIT 2 2BASES REFERENCES 1. WCAP-8403 (nonproprietary), "Power Distribution Control and LoadFollowing Procedures," Westinghouse Electric Corporation,September 1974.2.R. W. Miller et al., "Relaxation of Constant Axial Offset Control: FQSurveillance Technical Specification," WCAP-10217(NP), June 1983.3.FSAR, Chapter [15].UFSAR, Section 4.3.2.BAW10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. 222Enclosure 2, Volume 7, Rev. , Page 200 of 249Enclosure 2, Volume 7, Rev. , Page 200 of 249312.UFSAR,Chapter15. AFD (RAOC Methodology) B 3.2.3B Westinghouse STS B 3.2.3B-5 Rev. 4.0 1Revision XXX1SEQUOYAH UNIT 2 2 5 JUSTIFICATION FOR DEVIATIONS ITS 3.2.3 BASES, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The type of Methodology (Relaxed Axial Offset Control (RAOC)) and the Specification designator "B" are deleted since they are unnecessary (only one AFD Specification is used in the Sequoyah Nuclear (SQN) Plant ITS.) This information is provided in NUREG-1431, Rev. 4.0, to assist in indentifying the appropriate Specification to be used as a model for the plant specific ITS conversion, but serves no purpose in a plant specific implementation. In addition, the Constant Axial Offset Control (CAOC) methodology Specification (ISTS B 3.2.3A) is not used and is not shown.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS SR 3.2.3.1 Bases provides two options for controlling the Frequency of Surveillance Requirement. SQN is proposing to control the Surveillance Frequency under the Surveillance Frequency Control Program. Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 5. ISTS 3.2.3 Bases contains Figure B 3.2.3B-1. This Figure is located in the Sequoyah Nuclear Plant (SQN) COLR. Therefore, this figure is not included in the Bases for ITS 3.2.3.

6. Editorial changes made to enhance clarity/consistency.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.2.3, AXIAL FLUX DIFFERENCE (AFD) Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 4 ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER* ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until: a) Either the QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Within 2 hours: a) Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.

4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. __________________ *See Special Test Exception 3.10.2.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-12 Amendment No. 138, 223 be A02A03L0112 hours A04not within limit A05A05L03L02LCO 3.2.4 Applicability ACTION A, ACTION B ACTION A ACTION B ACTION A A03Add proposed Required Actions A.3, A.4, A.5, and A.6 and proposed ACTION B Page 1 of 6 or equal to A06after each QPTR determination M01 A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS ACTION: (Continued) b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until: a) Either the QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 within 30 minutes. 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. 4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until: a) Either the QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

April 21, 1997 SEQUOYAH - UNIT 1 3/4 2-13 Amendment No. 138, 223 A05L0112 hours A04not within limit L03A0412 hoursnot within limitL01L042 hours A05ACTION A, ACTION B ACTION A ACTION B ACTION A, ACTION B ACTION A ACTION B ACTION A Page 2 of 6 or equal to A06or equal to A06 A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS ACTION: (Continued) 2.Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. 3.Identify and correct the cause of the out of limit condition prior to increasingTHERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for l2 hours or until verified at 95% or greater RATED THERMAL POWER. d.With the indicated QUADRANT POWER TILT RATIO not confirmed as required bySurveillance Requirement 4.2.4.2, reduce THERMAL POWER to less than 75 percentRATED THERMAL POWER within 6 hours. e.With the QUADRANT POWER TILT RATIO not monitored as required by SurveillanceRequirement 4.2.4.1, reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: a.Calculating the ratio at least once per 7 days when the alarm is OPERABLE.b.Calculating the ratio at least once per 12 hours during steady state operation when thealarm is inoperable. 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from the 4 pairs of symmetric thimble locations or from performance of a full core map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. April 11, 2005 SEQUOYAH - UNIT 1 3/4 2-14 Amendment Nos. 135, 138, 301 L03L05L05L06Add proposed SR 3.2.4.1 Notes 1 and 2LA01LA01LA02SR 3.2.4.1 In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program L07Add proposed SR 3.2.4.2 Note SR 3.2.4.2 SR 3.2.4.2 Note SR 3.2.4.2 L08Page 3 of 6 A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER* ACTION: a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUARANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. 2. Within 2 hours either: a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.

4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exception 3.10.2.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-10 Amendment No. 130, 214 be A02A03L0112 hoursA04A05L03LCO 3.2.4 Applicability ACTION A, ACTION B ACTION A ACTION B ACTION A A03Add proposed Required Actions A.3, A.4, A.5, and A.6 and proposed ACTION B L02A05Page 4 of 6 or equal to A06not within limit after each QPTR determination M01 A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS ACTION: (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 within 30 minutes. 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. 4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

April 21, 1997 SEQUOYAH - UNIT 2 3/4 2-11 Amendment No. 130, 214 A05L0112 hours A04not within limit L03A0412 hoursnot within limitL01A05ACTION A, ACTION B ACTION A ACTION B ACTION A, ACTION B ACTION A ACTION B ACTION A L042 hours Page 5 of 6 or equal to A06or equal to A06 A01ITS ITS 3.2.4 POWER DISTRIBUTION LIMITS ACTION: (Continued) 2.Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. 3.Identify and correct the cause of the out of limit condition prior to increasing THERMALPOWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. d.With the indicated QUADRANT POWER TILT RATIO not confirmed as required bySurveillance Requirement 4.2.4.2, reduce THERMAL POWER to less than 75 percent RATED THERMAL POWER within 6 hours. e.With the QUADRANT POWER TILT RATIO not monitored as required by SurveillanceRequirement 4.2.4.1, reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: a.Calculating the ratio at least once per 7 days when the alarm is OPERABLE.b.Calculating the ratio at least once per 12 hours during steady state operation when thealarm is inoperable. 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from 4 pairs of symmetric thimble locations or from performance of a full core map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. April 11, 2005 SEQUOYAH - UNIT 2 3/4 2-12 Amendment No. 122, 130, 290 L03L05L05L06Add proposed SR 3.2.4.1 Notes 1 and 2LA01LA01LA02SR 3.2.4.1 In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program L07Add proposed SR 3.2.4.2 Note SR 3.2.4.2 SR 3.2.4.2 NoteSR 3.2.4.2 L08Page 6 of 6 DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 1 of 9 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.2.4 states "The QUADRANT POWER TILT RATIO shall not exceed 1.02." ITS LCO 3.2.4 states "The QPTR shall be 1.02. This changes the CTS by requiring the QPTR to be less than or equal to 1.02. This change is acceptable because nothing has changed. This is a presentation change for clarity. Stating that the QPTR shall be less than or equal to 1.02 is clearer than stating that it shall not exceed. This change is designated as an administrative change because it does not result in a technical change to the CTS. A03 CTS 3.2.4 Applicability contains a footnote (footnote *) that states "See Special Test Exceptions 3.10.2." ITS 3.2.4 Applicability does not contain this footnote. This changes the CTS by not including the footnote reference. The purpose of CTS 3.2.4 footnote

is to alert the user that a Special Test Exception exists which may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as an administrative change since it does not result in a technical change to the CTS. A04 CTS 3.2.4 ACTION a states "With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09." CTS 3.2.4 ACTION b states "With the QUADRANT POWER TILT RATIO determined to exceed 1.09 resulting from misalignment of either a shutdown or control rod." CTS 3.2.4 ACTION c states "With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod." ITS 3.2.4 ACTION A states "QPTR not within limit." This changes the CTS by specifying that action must be taken when the QPTR is not within limits. (See DOCS L02, L03, and L04 for changes to the compensatory measures.) The purpose of CTS 3.2.4 is to provide compensatory actions when the QPTR exceeds 1.02. ITS 3.2.4 continues to provide compensatory actions when the QPTR exceeds 1.02. This change is a presentation change. This change is designated as an administrative change since it does not result in technical changes to the CTS.

DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 2 of 9 A05 CTS 3.2.4 ACTION a.1.a) states that with QPTR greater than 1.02 and less than or equal to 1.09, calculate the QUADRANT POWER TILT RATIO at least once per hour until either QUADRANT POWER TILT RATIO is reduced to within its limit or THERMAL POWER is reduced to less than 50% of RTP. CTS 3.2.4 ACTION a.2.a) states within 2 hours, either QUADRANT POWER TILT RATIO is reduced to within its limit or reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. CTS 3.2.4 ACTION b.1.a) states that with QPTR greater than 1.09 due to misalignment of either a shutdown or control rod, calculate the QUADRANT POWER TILT RATIO at least once per hour until either QUADRANT POWER TILT RATIO is reduced to within its limit or THEMAL POWER is reduced to less than 50% of RTP. CTS 3.2.4 ACTION c.1.a) states that with QPTR greater than 1.09 due to causes other than the misalignment of either a shutdown or control rod, calculate the QUADRANT POWER TILT RATIO at least once per hour until either QUADRANT POWER TILT RATIO is reduced to within its limit or THERMAL POWER is reduced to less than 50% of RTP. ITS 3.2.4 does not contain a Required Action stating QPTR must be reduced to within its limit. This changes the CTS by not specifically stating that the restoration of QUADRANT POWER TILT RATIO is required. This change is acceptable because the technical requirements have not changed. Restoration of compliance with the LCO is always an available Required Action. The convention in the ITS is to not state such "restore" options explicitly unless it is the only action or is required for clarity. This change is designated as an administrative change since it does not result in technical changes to the CTS. A06 CTS 3.2.4 LCO APPLICABLITY is MODE 1 above 50% RTP. CTS 3.2.4 ACTION a.1.b, ACTION b.1.b and ACTION c.1.b state, in part, to calculate the QUADRANT POWER TILT RATIO at least once per hour until either QUADRANT POWER TILT RATIO is reduced to within limit, or THERMAL POWER is reduced to less than 50% of RTP. ITS 3.2.4 LCO APPLICABILITY is MODE 1 with THERMAL POWER >50% RTP. ITS 3.2.4 CONDITION B states that when the Required Action and associated Completion Time are not met to reduce THERMAL POWER to 50% RTP. This changes the CTS requirement of reducing power and exiting the MODE of APPLICABILITY to a value of < 50% RTP and allow stopping at a value of 50% RTP. This change is acceptable because the technical requirements have not changed. LCO 3.0.2 states that that when a Required Action to restore variables within limits is not met, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. In this case, both CTS and ITS require a reduction of power to exit the MODE of APPLICABILITY when compliance with the LCO is not met within the prescribed amount of time. Once the MODE of APPLICABILITY for LCO 3.2.4 is exited(>50%), the new power level(50%) is no longer controlled by this specification. This change is designated as an administrative change since it does not result in technical changes to CTS LCO 3.2.4.

DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 3 of 9 MORE RESTRICTIVE CHANGES M01 CTS 3.2.4 ACTION a.2.b states in part, within 2 hours, reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02. ITS 3.2.4 Required Action A.1 has a similar requirement to reduce THERMAL POWER 3% from RTP for each 1% of QPTR > 1.02. The Completion Time for ITS 3.2.4 Required Action A.1 is 2 hours after each QPTR determination. This changes the CTS by specifically requiring a power reduction, if applicable, after each QPTR determination. The purpose CTS 3.2.4 ACTION a.2.b is to commence a power level reduction to ensure that core power distributions that violate fuel design criteria are minimized. The maximum allowable power level initially determined by ITS 3.2.4 Required Action A.1 may be affected by subsequent determinations of QPTR. However, any increases in QPTR would require additional power reductions within 2 hours of each QPTR determination, if necessary to comply with the decreased maximum allowable power level. This change is designated as more restrictive because it adds required actions to the CTS.

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.2.4.1 states, in part, the QPTR shall be determined at least once per 7 days by calculating the ratio. CTS 4.2.4.2 states, in part, the QPTR shall be determined, at least once per 12 hours, by using the movable incore detectors. ITS SR 3.2.4.1 and SR 3.2.4.2 require similar Surveillances and specify the periodic Frequencies as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for these SRs and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 4 of 9 associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.2.4.2 states, in part, that the QPTR shall be determined to be within the limit by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from the 4 pairs of symmetric thimble locations or from performance of a full core map, is consistent with the indicated QUADRANT POWER TILT RATIO. ITS SR 3.2.4.2 requires verifying QPTR is within limit using the movable incore detectors. This changes the CTS by moving the procedural details for meeting the Surveillance to the Bases. The removal of these details, which are related to system design, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide protection of public health and safety. The ITS still retains the requirement that the QPTR is verified to be within the limits using the movable incore detectors. The details relating to system design do not need to appear in the specification in order for the requirement to apply. Additionally, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.2.4 ACTIONS a.1, b.1, and c.1 require calculating the QPTR at least once per hour. ITS 3.2.4 ACTION A (Required Action A.2 and associated Completion Time) require, in part, that when the QPTR is not within limit to determine QPTR once per 12 hours. This changes the CTS by requiring the determination of QPTR to be done once per 12 hours instead of once per hour. The purpose of CTS 3.2.4 ACTIONS a.1, b.1, and c.1 is to verify QPTR until it is brought to within limit or reactor power has been lowered to less than or equal to 50% RTP. This action is taken because with the QPTR not within limit, the core power distribution is not within the analyzed assumptions, and critical parameters such as FQ (X, Y, Z) and FH (X,Y) may not be within their limits. In addition to ITS 3.2.4 Required Action A.2 Completion Time the other Required Actions and associated Completion Times of Condition A are consistent with safe operation, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. In addition to reducing reactor power by greater than or equal to 3% for each 1% QPTR exceeds 1.02, ITS 3.2.4 requires a determination of QPTR once per 12 hours. Additionally, ITS 3.2.4 requires DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 5 of 9 measurement of FQ (X, Y, Z) and FH (X,Y) within 24 hours and every 7 days thereafter to verify that those parameters are within limit. Furthermore, ITS 3.2.4 requires the safety analyses to be reevaluated to ensure that the results remain valid. Assuming that these actions are successful, ITS 3.2.4 allows indefinite operation with QPTR out of its limit and allows the excore nuclear detectors to be normalized to eliminate the indicated QPTR. This ensures the core is operated within the safety analyses. This change is designated as less restrictive because less stringent Completion Times are being applied in the ITS than were applied in the CTS. L02 (Category 4 - Relaxation of Required Action) CTS 3.2.4 ACTION a.2.b) requires that when QPTR is in excess of 1.02 but less than or equal to 1.09, to reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. ITS 3.2.4 Required Action A.1 includes the requirement to reduce the THERMAL POWER, but does not include a requirement to reduce the Power Range Neutron Flux-High Trip Setpoints. This changes the CTS by eliminating the requirement to reduce the Power Range Neutron Flux-High Trip Setpoints. The purpose of CTS 3.2.4 ACTION a.2.b) is to reduce THERMAL POWER to increase the margin to the core power distribution limits. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while provided time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. With THERMAL POWER reduced by 3% from RTP for each 1% QPTR is greater than 1.02, further actions are not required to ensure that THERMAL POWER is not increased. Power increases are administratively prohibited by the Technical Specification while avoiding the risk of changing Reactor Trip System setpoints during operation. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L03 (Category 4 - Relaxation of Required Action) CTS 3.2.4 ACTION a.3 states "Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours." CTS 3.2.4 ACTION b.3 and b.4 contain the same compensatory actions as CTS ACTION a.3 but requires the QPTR to be within limits within 2 hours. CTS 3.2.4 ACTIONS a.4, b.4, and c.3 state "Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER." DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 6 of 9 ITS 3.2.4 Required Action A.3 requires performance of SR 3.2.1.1, SR 3.2.1.2, SR 3.2.1.3, SR 3.2.2.1, SR 3.2.2.2 within 24 hours after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 and once per 7 days thereafter. ITS 3.2.4 Required Action A.4 requires reevaluation of the safety analyses and confirmation that the results remain valid for duration of operation under this condition prior to increasing THERMAL POWER above the limit of Required Action A.1. ITS 3.2.4 Required Action A.5 requires normalization of excore detectors to restore QPTR to within limit prior to increasing THERMAL POWER above the limit of Required Action A.1. ITS 3.2.4 Required Action A.6 requires performance of SR 3.2.1.1, SR 3.2.1.2, SR 3.2.1.3, SR 3.2.2.1, SR 3.2.2.2 within 24 hours after achieving equilibrium conditions at RTP not to exceed 48 hours after increasing THERMAL POWER above the limit of Required Action A.1. Additionally, ITS 3.2.4 Required Action A.5 contains two Notes and ITS 3.2.4 Required Action A.6 contains one Note. ITS 3.2.4 Required Action A.5 Note 1 states "Perform Required Action A.5 only after Required Action A.4 is completed." ITS 3.2.4 Required Action A.5 Note 2 states "Required Action A.6 shall be completed whenever Required Action A.5 is performed." ITS 3.2.4 Required Action A.6 Note states "Perform Required Action A.6 only after Required Action A.5 is completed." Furthermore, ITS 3.2.4 ACTION B states that with a Required Action and associated Completion Time (of Condition A) not met, reduce THERMAL POWER to 50% RTP within 4 hours. This changes the CTS by eliminating requirements to be 50% RTP within a specified time of exceeding the LCO and substituting compensatory measures in ITS 3.2.4 ACTION A, which if not met, results in a reduction in power per ITS 3.2.4 ACTION B. The purpose of the CTS actions is to lower reactor power to less than 50% when QPTR is not within its limit and cannot be restored to within its limit within a reasonable time period. In addition, the Power Range Neutron Flux-High Trip setpoints are reduced to 55% to ensure that reactor power is not inadvertently increased without QPTR within its limit. This action is taken because with QPTR not within limit, the core power distribution is not within the analyzed assumptions, and critical parameters such as FQ (X, Y, Z) and FH (X,Y) may not be within their limits. A QPTR not within limit may not be an unacceptable condition if the critical core parameters such as FQ (X, Y, Z) and FH (X,Y) are within their limits. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while provided time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. ITS 3.2.4 requires measurement of FQ (X, Y, Z) and FH (X,Y) within 24 hours and every 7 days thereafter to verify that those parameters are within limit. In addition, ITS 3.2.4 requires the safety analyses to be reevaluated to ensure that the results remain valid. Assuming that these actions are successful, ITS 3.2.4 allows indefinite operation with QPTR out of its limit and allows the excore nuclear detectors to be normalized to eliminate the indicated QPTR. This ensures the core is operated within the safety analyses. This change is designated as less restrictive because less DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 7 of 9 stringent Required Actions are being applied in the ITS than were applied in the CTS. L04 (Category 3 - Relaxation of Completion Time) CTS 3.2.4 ACTION b.2, applies when QPTR is greater than 1.09 due to misalignment of either a shutdown or control rod, requires a THERMAL POWER reduction from RATED THERMAL POWER for each 1% of indicated QPTR in excess of 1.02 within 30 minutes. ITS 3.2.4 Required Action A.1 requires a THERMAL POWER reduction of 3% from RTP for each 1% QPTR exceeds 1.02 within 2 hours. This changes the CTS by allowing 2 hours to perform the required power reduction. The purpose of CTS 3.2.4 is to provide appropriate compensatory actions for QPTR greater than that assumed in the safety analyses. This change is acceptable because the completion Time is consistent with safe operation under the specified Condition, considering other indications available to the operator, a reasonable time for restoring compliance with the LCO, and the low probability of a DBA occurring during the restoration period. Under the ITS, a QPTR of 1.09 would require THERMAL POWER to be reduced to 79% RTP. This will provide sufficient thermal margin to account for the radial power distribution. In addition, the 2 hour time limit is consistent with the CTS time allowed when QPTR is > 1.02 but 1.09. This change is designated as less restrictive because additional time is allowed to decrease power than was allowed in the CTS. L05 (Category 4 - Relaxation of Required Action) CTS 3.2.4 ACTION d states "With the indicated QUADRANT POWER TILT RATIO not confirmed as required by Surveillance Requirement 4.2.4.2, reduce THERMAL POWER to less than 75 percent RATED THERMAL POWER within 6 hours." CTS 3.2.4 ACTION e states "With the QUADRANT POWER TILT RATIO not monitored as required by Surveillance Requirement 4.2.4.1, reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within the next 6 hours." ITS 3.2.4 does not contain these ACTIONS. This changes the CTS by not requiring RTP to be reduced to less than 75 percent, within 6 hours, when the QPTR is not confirmed and not requiring RTP to be reduced to less than 50 percent, within 6 hours, when the QPTR is not monitored. The purpose of CTS 3.2.4 ACTIONs d and e is to provide compensatory actions to take when Surveillance 4.2.4.1 has not been met or Surveillance 4.2.4.2 have not been performed. ITS 3.2.4 does not contain these ACTIONS since ITS SR 3.0.1 and SR 3.0.3 provide guidance on missed and not performed Surveillances. ITS SR 3.0.1 states, in part, that failure to meet a Surveillance is a failure to meet the LCO. Therefore, the compensatory actions for ITS LCO 3.2.4 would be entered. Additionally, ITS SR 3.0.1 states, in part, that failure to perform a Surveillance shall be failure to meet the LCO, but allows an exception as provided in SR 3.0.3. ITS SR 3.0.3 allows a delayed entry into the LCO to perform the Surveillance. If the Surveillance is not performed in this time period, then the LCO must be declared not met and the compensatory actions for ITS LCO 3.2.4 entered. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 8 of 9 L06 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 4.2.4.1.a states, in part, that the QPTR shall be determined to be within the limit by calculating the ratio at least once per 7 days. ITS SR 3.2.4.1 requires the same determination, but includes two Notes. ITS SR 3.2.4.1 Note 1 states when the input from one Power Range Neutron Flux channel is inoperable, the remaining three power range channels can be used for calculating QPTR as long as THERMAL POWER is less than or equal to 75% RTP. ITS SR 3.2.4.1 Note 2 states that SR 3.2.4.2 may be performed in lieu of this Surveillance. This changes the CTS by allowing use of three Power Range Neutron Flux channels for calculating the QPTR and by allowing the movable incore detectors to be used to determine QPTR instead of the excore detectors. The purpose of CTS 4.2.4.1.a is to periodically verify that QPTR is within limit. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are sufficient for verification that the parameters meet the LCO. When one or more Power Range Neutron Flux channels are inoperable, tilt monitoring becomes degraded. With only one Power Range Neutron Flux channel inoperable, QPTR can still be verified by calculation as long as three Power Range Neutron Flux channels are OPERABLE and THERMAL POWER is less than or equal to 75% RTP. The movable incore detector system provides a more accurate indication of QPTR than the excore detectors. In fact, the movable incore detector system is used to calibrate the excore detectors. Therefore, allowing the use of the movable incore detector system or excore detector is appropriate. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. L07 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 4.2.4.1.a states that the QPTR shall be determined to be within the limit by calculating the ratio at least once per 7 days when the alarm is OPERABLE. CTS 4.2.4.1.b states that the QPTR shall be determined to be within the limit by calculating the ratio at least once per 12 hours during steady state operation when the alarm is inoperable. ITS SR 3.2.4.1 requires verification that the QPTR is within limits every 7 days. This changes the CTS by eliminating the requirement to verify the QPTR more frequently when the QPTR alarm is inoperable. The purpose of CTS 4.2.4.1.a and 4.2.4.1.b is to periodically verify that the QPTR is within limit. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the frequency of QPTR verification when the QPTR alarm is inoperable is unnecessary as inoperability of the alarm does not increase the probability that the QPTR is outside its limit. The QPTR alarm is for indication only. It use is not credited in any of the safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L08 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 4.2.4.2 states, in part, that the QPTR shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors. ITS DISCUSSION OF CHANGES ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 9 of 9 SR 3.2.4.2 requires determination of the QPTR by use of the movable incore detectors. Additionally, ITS SR 3.2.4.2 contains a Note which states "Not required to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP." This changes the CTS by not requiring the Surveillance to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable. The purpose of CTS 4.2.4.2 is to verify that the QPTR is within limit using the movable incore detectors. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. When one or more Power Range Neutron Flux channels are inoperable, tilt monitoring becomes degraded. Therefore, the movable incore detector system provides a more accurate indication of QPTR than the excore detectors. The ITS SR 3.2.4.2 allowance, for not requiring performance of the Surveillance for 12 hours after input when one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP, is required to allow time for the movable incore detectors to perform the initial measurement of the QPTR before the Surveillance is declared not met. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) QPTR 3.2.4 Westinghouse STS 3.2.4-1 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 1 23.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. QPTR not within limit. A.1 Reduce THERMAL POWER 3% from RTP for each 1% of QPTR > 1.00. AND A.2 Determine QPTR. AND A.3 Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

AND 2 hours after each QPTR determination

Once per 12 hours

24 hours after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter 3.2.4 Applicability ACTION a, ACTION b, ACTION c DOC M01 121.02SR 3.2.1.3, SR 3.2.2.1 and SR 3.2.2.2.DOC L03 QPTR 3.2.4 Westinghouse STS 3.2.4-2 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 1 2ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.4 Reevaluate safety analyses and confirm results remain valid for duration of operation under this condition. AND A.5 -------------NOTES------------- 1. Perform Required Action A.5 only after Required Action A.4 is completed. 2. Required Action A.6 shall be completed whenever Required Action A.5 is performed. ------------------------------------- Normalize excore detectors to restore QPTR to within limit. AND A.6 ---------------NOTE-------------- Perform Required Action A.6 only after Required Action A.5 is completed. -------------------------------------

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1. Prior to increasing THERMAL POWER above the limit of Required Action A.1

Prior to increasing THERMAL POWER above the limit of Required Action A.1

Within 24 hours after achieving equilibrium conditions at RTP not to exceed 48 hours after increasing THERMAL POWER above the limit of Required Action A.1 DOC L03 1SR 3.2.1.3, SR 3.2.2.1 and SR 3.2.2.2.DOC L03 DOC L03 QPTR 3.2.4 Westinghouse STS 3.2.4-3 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 1 2ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time not met. B.1 Reduce THERMAL POWER to 50% RTP. 4 hours

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 ------------------------------NOTES-----------------------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance. --------------------------------------------------------------------- Verify QPTR is within limit by calculation.

 [ 7 days  OR In accordance with the Surveillance Frequency Control Program ]  ACTION a, ACTION b, ACTION c 4.2.4.1 DOC L06 33 QPTR 3.2.4   Westinghouse STS 3.2.4-4 Rev. 4.0   CTS Amendment XXX SEQUOYAH UNIT 1 2SURVEILLANCE REQUIREMENTS  (continued)  SURVEILLANCE   FREQUENCY   SR  3.2.4.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP. ---------------------------------------------------------------------

Verify QPTR is within limit using the movable incore detectors.

[ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 334.2.4.2, DOC L08 QPTR 3.2.4 Westinghouse STS 3.2.4-1 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 2 23.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. QPTR not within limit. A.1 Reduce THERMAL POWER 3% from RTP for each 1% of QPTR > 1.00. AND A.2 Determine QPTR. AND A.3 Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

AND 2 hours after each QPTR determination

Once per 12 hours

24 hours after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter 3.2.4 Applicability ACTION a, ACTION b, ACTION c DOC M01 121.02SR 3.2.1.3, SR 3.2.2.1 and SR 3.2.2.2.DOC L03 QPTR 3.2.4 Westinghouse STS 3.2.4-2 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 2 2ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.4 Reevaluate safety analyses and confirm results remain valid for duration of operation under this condition. AND A.5 -------------NOTES------------- 1. Perform Required Action A.5 only after Required Action A.4 is completed. 2. Required Action A.6 shall be completed whenever Required Action A.5 is performed. ------------------------------------- Normalize excore detectors to restore QPTR to within limit. AND A.6 ---------------NOTE-------------- Perform Required Action A.6 only after Required Action A.5 is completed. -------------------------------------

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1. Prior to increasing THERMAL POWER above the limit of Required Action A.1

Prior to increasing THERMAL POWER above the limit of Required Action A.1

Within 24 hours after achieving equilibrium conditions at RTP not to exceed 48 hours after increasing THERMAL POWER above the limit of Required Action A.1 DOC L03 1SR 3.2.1.3, SR 3.2.2.1 and SR 3.2.2.2.DOC L03 DOC L03 QPTR 3.2.4 Westinghouse STS 3.2.4-3 Rev. 4.0 CTS Amendment XXX SEQUOYAH UNIT 2 2ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time not met. B.1 Reduce THERMAL POWER to 50% RTP. 4 hours

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 ------------------------------NOTES-----------------------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance. --------------------------------------------------------------------- Verify QPTR is within limit by calculation.

 [ 7 days  OR In accordance with the Surveillance Frequency Control Program ]  ACTION a, ACTION b, ACTION c 4.2.4.1 DOC L06 33 QPTR 3.2.4   Westinghouse STS 3.2.4-4 Rev. 4.0   CTS Amendment XXX SEQUOYAH UNIT 2 2SURVEILLANCE REQUIREMENTS  (continued)  SURVEILLANCE   FREQUENCY   SR  3.2.4.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP. ---------------------------------------------------------------------

Verify QPTR is within limit using the movable incore detectors.

[ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 334.2.4.2, DOC L08 JUSTIFICATION FOR DEVIATIONS ITS 3.2.4, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made to be consistent with changes made to Specification 3.2.1 and 3.2.2. 2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS SR 3.2.4.1 and SR 3.2.4.2 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) QPTR B 3.2.4 Westinghouse STS B 3.2.4-1Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR) BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation. The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.6, "Control Rod Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY design criteria: ANALYSES a.During a large break loss of coolant accident, the peak claddingtemperature must not exceed 2200°F (Ref. 1),b.During a loss of forced reactor coolant flow accident, there must be atleast 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition,c.During an ejected rod accident, the energy deposition to the fuelmust not exceed 280 cal/gm (Ref. 2), andd.The control rods must be capable of shutting down the reactor with aminimum required SDM with the highest worth control rod stuck fullywithdrawn (Ref. 3).The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor (FQ(Z)), the Nuclear Enthalpy Rise Hot Channel Factor )F(HN, and control bank insertion are established to preclude core power distributions that exceed the safety analyses limits. The QPTR limits ensure that HNF and FQ(Z) remain below their limiting values by preventing an undetected change in the gross radial power distribution. X, Y, (FH(X, Y)) X, Y, FH(X, Y) 111 QPTR B 3.2.4 Westinghouse STS B 3.2.4-2Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 BASES APPLICABLE SAFETY ANALYSES (continued) In MODE 1, the HNF and FQ(Z) limits must be maintained to preclude core power distributions from exceeding design limits assumed in the safety analyses. The QPTR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ(Z) and )F(HN is possibly challenged. APPLICABILITY The QPTR limit must be maintained in MODE 1 with THERMAL POWER > 50% RTP to prevent core power distributions from exceeding the design limits. Applicability in MODE 1 50% RTP and in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the HNF and FQ(Z) LCOs still apply, but allow progressively higher peaking factors at 50% RTP or lower. ACTIONSA.1 With the QPTR exceeding its limit, a power level reduction of 3% RTP for each 1% by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours allows sufficient time to identify the cause and correct the tilt. Note that the power reduction itself may cause a change in the tilted condition. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of QPTR. Increases in QPTR would require power reduction within 2 hours of QPTR determination, if necessary to comply with the decreased maximum allowable power level. Decreases in QPTR would allow increasing the maximum allowable power level and increasing power up to this revised limit. X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) 1111.021 QPTR B 3.2.4 Westinghouse STSB 3.2.4-3Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 BASES ACTIONS (continued) A.2 After completion of Required Action A.1, the QPTR alarm may still be in its alarmed state. As such, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours thereafter. A 12 hour Completion Time is sufficient because any additional change in QPTR would be relatively slow. A.3 The peaking factors FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are of primary importance in ensuring that the power distribution remains consistent with the initial conditions used in the safety analyses. Performing SRs on HNF and FQ(Z) within the Completion Time of 24 hours after achieving equilibrium conditions from a Thermal Power reduction per Required Action A.1 ensures that these primary indicators of power distribution are within their respective limits. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to support flux mapping. A Completion Time of 24 hours after achieving equilibrium conditions from Thermal Power reduction per Required Action A.1 takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perform a flux map. If these peaking factors are not within their limits, the Required Actions of these Surveillances provide an appropriate response for the abnormal condition. If the QPTR remains above its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate HNF and FQ(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit. A.4 Although HNF and FQ(Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) 1111 the applicable LCOs 3 QPTR B 3.2.4 Westinghouse STS B 3.2.4-4Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 BASES ACTIONS (continued) malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the limit of Required Action A.1, the reactor core conditions are consistent with the assumptions in the safety analyses. A.5 If the QPTR has exceeded the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are normalized to restore QPTR to within limits prior to increasing THERMAL POWER to above the limit of Required Action A.1. Normalization is accomplished in such a manner that the indicated QPTR following normalization is near 1.00. This is done to detect any subsequent significant changes in QPTR. Required Action A.5 is modified by two Notes. Note 1 states that the QPTR is not restored to within limits until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., Required Action A.4). Note 2 states that if Required Action A.5 is performed, then Required Action A.6 shall be performed. Required Action A.5 normalizes the excore detectors to restore QPTR to within limits, which restores compliance with LCO 3.2.4. Thus, Note 2 prevents exiting the Actions prior to completing flux mapping to verify peaking factors, per Required Action A.6. These Notes are intended to prevent any ambiguity about the required sequence of actions. A.6 Once the flux tilt is restored to within limits (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution is consistent with the safety analysis assumptions, Required Action A.6 requires verification 11.02is still exceedingshall be shall not be by excore detector normalization 266 QPTR B 3.2.4 Westinghouse STS B 3.2.4-5 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 BASES

ACTIONS (continued) that FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are within their specified limits within 24 hours of achieving equilibrium conditions at RTP. As an added precaution, if the core power does not reach equilibrium conditions at RTP within 24 hours, but is increased slowly, then the peaking factor surveillances must be performed within 48 hours after increasing THERMAL POWER above the limit of Required Action A.1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A.1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time. Required Action A.6 is modified by a Note that states that the peaking factor surveillances may only be done after the excore detectors have been normalized to restore QPTR to within limits (i.e., Required Action A.5). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.

B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems. SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1. This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. [ The Frequency of 7 days takes into account other information and alarms available to the operator in the control room. 41X, Y, FH(X, Y) 1 QPTR B 3.2.4 Westinghouse STS B 3.2.4-6 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 1 BASES SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP. With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. [ Performing SR 3.2.4.2 at a Frequency of 12 hours provides an accurate alternative means for ensuring that any tilt remains within its limits. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

4544544QPTR6 QPTR B 3.2.4 Westinghouse STS B 3.2.4-7 Rev. 4.0 2Revision XXXSEQUOYAH UNIT 1 BASES SURVEILLANCE REQUIREMENTS (continued) For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8 for three and four loop cores. The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. Therefore, incore monitoring of QPTR can be used to confirm that QPTR is within limits.

With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data. REFERENCES 1. 10 CFR 50.46.

2. Regulatory Guide 1.77, Rev [0], May 1974. 3. 10 CFR 50, Appendix A, GDC 26. 22 QPTR B 3.2.4 Westinghouse STS B 3.2.4-1 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 B 3.2 POWER DISTRIBUTION LIMITS

B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation. The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.6, "Control Rod Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY design criteria: ANALYSES a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200°F (Ref. 1), b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition, c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2), and d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3). The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor (FQ(Z)), the Nuclear Enthalpy Rise Hot Channel Factor )F(HN, and control bank insertion are established to preclude core power distributions that exceed the safety analyses limits. The QPTR limits ensure that HNF and FQ(Z) remain below their limiting values by preventing an undetected change in the gross radial power distribution. X, Y, (FH(X, Y)) X, Y, FH(X, Y) 111 QPTR B 3.2.4 Westinghouse STS B 3.2.4-2 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 BASES

APPLICABLE SAFETY ANALYSES (continued) In MODE 1, the HNF and FQ(Z) limits must be maintained to preclude core power distributions from exceeding design limits assumed in the safety analyses. The QPTR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ(Z) and )F(HN is possibly challenged. APPLICABILITY The QPTR limit must be maintained in MODE 1 with THERMAL POWER > 50% RTP to prevent core power distributions from exceeding the design limits. Applicability in MODE 1 50% RTP and in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the HNF and FQ(Z) LCOs still apply, but allow progressively higher peaking factors at 50% RTP or lower. ACTIONS A.1 With the QPTR exceeding its limit, a power level reduction of 3% RTP for each 1% by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours allows sufficient time to identify the cause and correct the tilt. Note that the power reduction itself may cause a change in the tilted condition. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of QPTR. Increases in QPTR would require power reduction within 2 hours of QPTR determination, if necessary to comply with the decreased maximum allowable power level. Decreases in QPTR would allow increasing the maximum allowable power level and increasing power up to this revised limit. X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) 1111.021 QPTR B 3.2.4 Westinghouse STSB 3.2.4-3Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 BASES ACTIONS (continued) A.2 After completion of Required Action A.1, the QPTR alarm may still be in its alarmed state. As such, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours thereafter. A 12 hour Completion Time is sufficient because any additional change in QPTR would be relatively slow. A.3 The peaking factors FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are of primary importance in ensuring that the power distribution remains consistent with the initial conditions used in the safety analyses. Performing SRs on HNF and FQ(Z) within the Completion Time of 24 hours after achieving equilibrium conditions from a Thermal Power reduction per Required Action A.1 ensures that these primary indicators of power distribution are within their respective limits. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to support flux mapping. A Completion Time of 24 hours after achieving equilibrium conditions from Thermal Power reduction per Required Action A.1 takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perform a flux map. If these peaking factors are not within their limits, the Required Actions of these Surveillances provide an appropriate response for the abnormal condition. If the QPTR remains above its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate HNF and FQ(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit. A.4 Although HNF and FQ(Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) X, Y, FH(X, Y) 1111 the applicable LCOs 3 QPTR B 3.2.4 Westinghouse STS B 3.2.4-4 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 BASES

ACTIONS (continued) malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the limit of Required Action A.1, the reactor core conditions are consistent with the assumptions in the safety analyses. A.5 If the QPTR has exceeded the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are normalized to restore QPTR to within limits prior to increasing THERMAL POWER to above the limit of Required Action A.1. Normalization is accomplished in such a manner that the indicated QPTR following normalization is near 1.00. This is done to detect any subsequent significant changes in QPTR. Required Action A.5 is modified by two Notes. Note 1 states that the QPTR is not restored to within limits until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., Required Action A.4). Note 2 states that if Required Action A.5 is performed, then Required Action A.6 shall be performed. Required Action A.5 normalizes the excore detectors to restore QPTR to within limits, which restores compliance with LCO 3.2.4. Thus, Note 2 prevents exiting the Actions prior to completing flux mapping to verify peaking factors, per Required Action A.6. These Notes are intended to prevent any ambiguity about the required sequence of actions. A.6 Once the flux tilt is restored to within limits (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution is consistent with the safety analysis assumptions, Required Action A.6 requires verification 11.02is still exceedingshall be shall not be by excore detector normalization 266 QPTR B 3.2.4 Westinghouse STS B 3.2.4-5 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 BASES

ACTIONS (continued) that FQ(Z), as approximated by )Z(FCQ and )Z(FWQ, and HNF are within their specified limits within 24 hours of achieving equilibrium conditions at RTP. As an added precaution, if the core power does not reach equilibrium conditions at RTP within 24 hours, but is increased slowly, then the peaking factor surveillances must be performed within 48 hours after increasing THERMAL POWER above the limit of Required Action A.1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A.1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time. Required Action A.6 is modified by a Note that states that the peaking factor surveillances may only be done after the excore detectors have been normalized to restore QPTR to within limits (i.e., Required Action A.5). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.

B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems. SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1. This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. [ The Frequency of 7 days takes into account other information and alarms available to the operator in the control room. 41X, Y, FH(X, Y) 1 QPTR B 3.2.4 Westinghouse STS B 3.2.4-6 Rev. 4.0 2Revision XXX SEQUOYAH UNIT 2 BASES SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP. With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. [ Performing SR 3.2.4.2 at a Frequency of 12 hours provides an accurate alternative means for ensuring that any tilt remains within its limits. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

4544544QPTR6 QPTR B 3.2.4 Westinghouse STS B 3.2.4-7 Rev. 4.0 2Revision XXXSEQUOYAH UNIT 2 BASES SURVEILLANCE REQUIREMENTS (continued) For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8 for three and four loop cores. The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. Therefore, incore monitoring of QPTR can be used to confirm that QPTR is within limits.

With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data. REFERENCES 1. 10 CFR 50.46.

2. Regulatory Guide 1.77, Rev [0], May 1974. 3. 10 CFR 50, Appendix A, GDC 26. 22 JUSTIFICATION FOR DEVIATIONS ITS 3.2.4 BASES, QUADRANT POWER TILT RATIO (QPTR) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made to be consistent with changes made to the Specification.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. ISTS 3.2.4 Bases Required Action A.3 refers to the Required Actions of the referenced Surveillances. There are no Required Actions in the ITS 3.2.1 or ITS 3.2.2 Surveillances. This reference has been corrected to refer to the Required Actions of the applicable LCOs. 4. ISTS SR 3.2.4.1 and SR 3.2.4.2 Bases provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Additionally, the Frequency description which is being removed will be included in the Surveillance Frequency Control Program. 5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 6. Typographical/grammatical error corrected. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.2.4, QUADRANT POWER TILT RATIO Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ENCLOSURE 2 VOLUME 9 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.4 REACTOR COOLANT SYSTEM (RCS)

Revision 0 LIST OF ATTACHMENTS 1. ITS 3.4.1 - RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits 2. ITS 3.4.2 - RCS Minimum Temperature for Criticality

3. ITS 3.4.3 - RCS Pressure and Temperature Limits 4. ITS 3.4.4 - RCS Loops - MODES 1 and 2

5. ITS 3.4.5 - RCS Loops - MODE 3

6. ITS 3.4.6 - RCS Loops - MODE 4 7. ITS 3.4.7 - RCS Loops - MODE 5, Loops Filled 8. ITS 3.4.8 - RCS Loops - MODE 5, Loops Not Filled 9. ITS 3.4.9 - Pressurizer 10. ITS 3.4.10 - Pressurizer Safety Valves 11. ITS 3.4.11 - Pressurizer Power Operated Relief Valves

12. ITS 3.4.12 - Low Temperature Overpressure Protection System 13. ITS 3.4.13 - RCS Operational LEAKAGE

14. ITS 3.4.14 - RCS Pressure Isolation Valve Leakage 15. ITS 3.4.15 - RCS Leakage Detection Instrumentation 16. ITS 3.4.16 - RCS Specific Activity 17. ITS 3.4.17 - Steam Generator (SG) Tube Integrity 18. ISTS Not Adopted ATTACHMENT 1 ITS 3.4.1, RCS Pressure, Temperature, and Flow Departure From Nucleate Boiling (DNB) Limits Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

°

DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS

INSERT 1 M01 CTS Figure 3.2-1 Flow vs. Power for 4 Loops in Operation provides the acceptable Minimum Measured Reactor Coolant System Flow Rate (measured in gallons per minute (gpm)) based on the Thermal Power Fraction (% of Rated Thermal Power (RTP)). The Acceptable Operation Region covers flows from 100% RTP (minimum RCS flow rate 378,400 gpm) to 90% RTP (minimum RCS flow rate 359,480 gpm). ITS LCO 3.4.1 does not include a figure for a minimum acceptable RCS flow rate. ITS LCO 3.4.1 requires RCS total flow rate to be greater than or equal to the required flow for 100% RTP. This changes the CTS by not allowing the RCS total flow rate to be < 378,400 in MODE 1. The RCS flow rate normally remains constant during an operational fuel cycle with all pumps running. The minimum RCS flow limit corresponds to that assumed for Departure from Nucleate Boiling analyses. Therefore, this change is acceptable because it ensures the RCS flow rate assumed in the minimum departure from nucleate boiling ratio will be met for each transient analyzed. This change is designated as more restrictive because plant operations are more limited by the ITS requirements than the CTS requirements. DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

(Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, or ISI Program)

DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS (Type 6 - Removal of Cycle-Specific Parameter Limits from the Technical Specifications to the Core Operating Limits Report)

(Category 3 - Relaxation of Completion Time)

DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

;

INSERT 1

;

INSERT 1

JUSTIFICATION FOR DEVIATIONS ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1 INSERT 2

INSERT 3

INSERT 4

INSERT 1 INSERT 2

INSERT 3

INSERT 4

JUSTIFICATION FOR DEVIATIONS ITS 3.4.1 Bases, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS ATTACHMENT 2 ITS 3.4.2, RCS Minimum Temperature for Criticality Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs) °

°° ° °

DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY

°

DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY

(Category 7 - Relaxation Of Surveillance Frequency)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.2 BASES, RCS MINIMUM TEMPERATURE FOR CRITICALITY

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY ATTACHMENT 3 ITS 3.4.3, RCS Pressure and Temperature (P/T) Limits Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs) <

DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.3 BASES, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ATTACHMENT 4 ITS 3.4.4, RCS Loops - MODES 1 and 2 Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 and 2

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 and 2

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

(Category 3 - Relaxation of Completion Time) DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 and 2

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, RCS LOOPS - MODES 1 and 2

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1

INSERT 2

INSERT 1

INSERT 2

JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, RCS LOOPS - MODES 1 and 2

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.4, RCS LOOPS - MODES 1 and 2 ATTACHMENT 5 ITS 3.4.5, RCS Loops - MODE 3 Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

°

DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 (Type 1 - Removing Details of System Design and System Description, Including Design Limits)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

(Category 7 - Relaxation Of Surveillance Frequency)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.5, RCS LOOPS - MODE 3

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.5 Bases, RCS LOOPS - MODE 3

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.5, RCS LOOPS - MODE 3 ATTACHMENT 6 ITS 3.4.6, RCS LOOPS - MODE 4 Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

°

DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4

(Type 1 - Removing Details of System Design and System Description, Including Design Limits)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4

(Category 4 - Relaxation of Required Action)

(Category 7 - Relaxation of Surveillance Frequency)

DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.6, RCS Loops - MODE 4

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.6 Bases, RCS LOOPS - MODE 4

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.6, RCS LOOPS - MODE 4 ATTACHMENT 7 ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

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DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS FILLED

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS FILLED

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

(Category 1 - Relaxation of LCO Requirements)

DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS FILLED

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.7 Bases, RCS LOOPS - MODE 5, LOOPS FILLED Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED ATTACHMENT 8 ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

°

DISCUSSION OF CHANGES ITS 3.4.8, MODE 5, RCS LOOPS NOT FILLED

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

DISCUSSION OF CHANGES ITS 3.4.8, MODE 5, RCS LOOPS NOT FILLED

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.8 Bases, RCS LOOPS - MODE 5, LOOPS NOT FILLED

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED ATTACHMENT 9 ITS 3.4.9, PRESSURIZER Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

(Category 5 - Deletion of Surveillance Requirement)

DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER (Category 7 - Relaxation of Surveillance Frequency)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.9, PRESSURIZER

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1

JUSTIFICATION FOR DEVIATIONS ITS 3.4.9 Bases, PRESSURIZER

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.9, PRESSURIZER ATTACHMENT 10 ITS 3.4.10, PRESSURIZER SAFETY VALVES Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

+/-

DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES

DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

(Category 2 - Relaxation of Applicability)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

°

JUSTIFICATION FOR DEVIATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

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,

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JUSTIFICATION FOR DEVIATIONS ITS 3.4.10 Bases, PRESSURIZER SAFETY VALVES Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES ATTACHMENT 11 ITS 3.4.11, PRESSURIZER POWER OPERATED RELEIF VALVES (PORVs) Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

+/-

DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

(Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs)

(Category 3 - Relaxation of Completion Time)

(Category 7 - Relaxation of Surveillance Frequency)

DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) (Category 4 - Relaxation of Required Action)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs)

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.11 BASES, PRESSURIZER POWER OPERATED VALVES (PORVS)

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) ATTACHMENT 12 ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

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DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM

(Type 1 - Removing Details of System Design and System Description, Including Design Limits)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

INSERT 1

INSERT 2

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JUSTIFICATION FOR DEVIATIONS ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1

INSERT 2

INSERT 3

INSERT 4

INSERT 1

INSERT 2

INSERT 3

INSERT 4

JUSTIFICATION FOR DEVIATIONS ITS 3.4.12 Bases, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM ATTACHMENT 13 ITS 3.4.13, RCS OPERATIONAL LEAKAGE Current Technical Specification (CTS) Markupand Discussion of Changes (DOCs)

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DISCUSSION OF CHANGES ITS 3.4.13, RCS OPERATIONAL LEAKAGE

DISCUSSION OF CHANGES ITS 3.4.13, RCS OPERATIONAL LEAKAGE (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

333

22

333

22 JUSTIFICATION FOR DEVIATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

1 INSERT 1

INSERT 2

1 INSERT 1

INSERT 2

JUSTIFICATION FOR DEVIATIONS ITS 3.4.13 Bases, RCS OPERATIONAL LEAKAGE

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE ATTACHMENT 14 ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

(Type 1 - Removing Details of System Design and System Description, Including Design Limits)

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

(Category 7 - Relaxation of Surveillance Frequency)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1

JUSTIFICATION FOR DEVIATIONS ITS 3.4.14 BASES, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE ATTACHMENT 15 ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

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DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION

DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

(Type 1 - Removing Details of System Design and System Description, Including Design Limits)

DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION

(Category 4 - Relaxation of Required Action)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.15 BASES, RCS LEAKAGE DETECTION INSTRUMENTATION

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION ATTACHMENT 16 ITS 3.4.16, RCS SPECIFIC ACTIVITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

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DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT (µCi/gm)PERCENT OF RATED THERMAL POWERFIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit VersusPercent of RATED THERMAL POWER with the Primary Coolant SpecificActivity > 0.35 µCi/gram Dose Equivalent I-131 °°°

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DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT (µCi/gm)PERCENT OF RATED THERMAL POWERFIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit VersusPercent of RATED THERMAL POWER with the Primary Coolant SpecificActivity > 0.35 µCi/gram Dose Equivalent I-131 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY

DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY

(Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program)

DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY

(Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)

(Category 1 - Relaxation of LCO Requirements)

DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

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JUSTIFICATION FOR DEVIATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

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JUSTIFICATION FOR DEVIATIONS ITS 3.4.16 BASES, RCS SPECIFIC ACTIVITY

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY ATTACHMENT 17 ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Pages 3/4 4-7 through 3/4 4-12 intentionally deleted.

Pages 3/4 4-11 through 3/4 4-16 are intentionally deleted

DISCUSSION OF CHANGES ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

INSERT 1

JUSTIFICATION FOR DEVIATIONS ITS 3.4.17 BASES, STEAM GENERATOR (SG) TUBE INTEGRITY

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 18 Improved Standard Technical Specifications (ISTS) Not Adopted in the Sequoyah ITS ISTS 3.4.17, RCS LOOP ISOLATION VALVES

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.17, RCS LOOP ISOLATION VALVES Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.17 BASES, RCS LOOP ISOLATION VALVES

ISTS 3.4.18, RCS ISOLATION LOOP STARTUP Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ISTS 3.4.18, RCS ISOLATION LOOP STARTUP Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

1

1 JUSTIFICATION FOR DEVIATIONS ITS 3.4.18 BASES, RCS ISOLATION LOOP STARTUP

ISTS 3.4.19, RCS LOOPS - TEST EXCEPTIONS Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ISTS 3.4.19, RCS LOOPS - TEST EXCEPTIONS Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

JUSTIFICATION FOR DEVIATIONS ITS 3.4.19 BASES, RCS LOOPS - TEST EXCEPTIONS

ENCLOSURE 2 VOLUME 10 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) Revision 0 LIST OF ATTACHMENTS 1. ITS 3.5.1, - Accumulators 2. ITS 3.5.2, - ECCS - Operating

3. ITS 3.5.3 - ECCS - Shutdown 4. ITS 3.5.4 - Refueling Water Storage Tank (RWST) 5. ITS 3.5.5 - Seal Injection Flow

6. ISTS Not Adopted ATTACHMENT 1 ITS 3.5.1, ACCUMULATORS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01ITS ITS 3.5.13/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:

a. The isolation valve open, b. A contained borated water volume of between 7615 and 7960 gallons of borated water,

c. Between 2400 and 2700 ppm of boron,

d. A nitrogen cover-pressure of between 624 and 668 psig, and e. Power removed from isolation valve when RCS pressure is above 2000 psig. APPLICABILITY: MODES 1, 2 and 3.* ACTION: a. With one cold leg injection accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours.

b. With one cold leg injection accumulator inoperable due to the boron concentration not within limits, restore boron concentration to within limits within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours.

____________ *Pressurizer pressure above 1000 psig.

June 18, 2004 SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 124, 140, 147, 192, 262, 291 Page 1 of 4 LCO 3.5.1 Four ECCS SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.4 SR 3.5.1.3 SR 3.5.1.5 Applicability Applicability Add proposed ACTION D A03ACTION B ACTION C ACTION A ACTION C A02RCS A04RCS A04RCS A04A05greater than or equal to A01ITS ITS 3.5.1EMERGENCY CORE COOLING SYSTEMS (ECCS) SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: a. At least once per 12 hours by:

1. Verifying the contained borated water volume and nitrogen cover-pressure in each cold leg injection accumulator, and

2. Verifying that each cold leg injection accumulator isolation valve is fully open.

b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume, that is not the result of addition from the refueling water storage tank, # by verifying the boron concentration of the cold leg injection accumulator solution. c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is removed.

    # Only required to be performed for affected accumulators that experienced volume increases.

December 27, 1994 SEQUOYAH - UNIT 1 3/4 5-2 Amendment Nos. 12, 124, 147,192 Page 2 of 4 SR 3.5.1.2 SR 3.5.1.3 SR 3.5.1.1 SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.3 LA01In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 SR 3.5.1.5 SR 3.5.1.4 LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program A05greater than or equal to A01ITS ITS 3.5.13/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS

LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with: a. The isolation valve open,

b. A contained borated water volume of between 7615 and 7960 gallons of borated water,

c. Between 2400 and 2700 ppm of boron,

d. A nitrogen cover-pressure of between 624 and 668 psig, and

e. Power removed from isolation valve when RCS pressure is above 2000 psig.

APPLICABILITY: MODES 1, 2 and 3.* ACTION:

a. With one cold leg injection accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours.

b. With one cold leg injection accumulator inoperable due to the boron concentration not within limits, restore boron concentration to within limits within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours.

Pressurizer pressure above 1000 psig.

June 18. 2004 SEQUOYAH - UNIT 2 3/4 5-1 Amendment No. 113, 131, 133, 141, 184, 253, 281Four ECCS LCO 3.5.1 SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.4 SR 3.5.1.3 SR 3.5.1.5 Applicability Add proposed ACTION D A03Applicability Page 3 of 4 ACTION B ACTION C ACTION A ACTION C A02RCSA04RCSA04RCSA04A05greater than or equal A01ITS ITS 3.5.1EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours by:

1. Verifying the contained borated water volume and nitrogen cover-pressure each cold leg injection accumulator, and

2. Verifying that each cold leg injection accumulator isolation valve is fully open.

b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume, that is not the result of addition from the refueling water storage tank,# by verifying the boron concentration of the cold leg injection accumulator solution.

c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is removed.

______________________ # Only required to be performed for affected accumulators that experience volume increases.

December 27, 1994 SEQUOYAH - UNIT 2 3/4 5-2 Amendment No. 113, 133, 184 LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.3 SR 3.5.1.2 SR 3.5.1.3 SR 3.5.1.1 SR 3.5.1.4 SR 3.5.1.5 SR 3.5.1.4 Page 4 of 4 A05greater than or equal DISCUSSION OF CHANGES ITS 3.5.1, ACCUMULATORS Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.5.1.1 requires "each" cold leg injection accumulator to be OPERABLE. ITS LCO 3.5.1 requires "four" ECCS accumulators to be OPERABLE. This changes the CTS by specifying the exact number of ECCS accumulators required to be OPERABLE. The change is acceptable because the total number of ECCS accumulators in each unit at SQN is four. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 3.5.1.1 does not contain a specific ACTION for two or more accumulators inoperable. With two or more accumulators inoperable, CTS 3.0.3 would be entered. ITS 3.5.1 ACTION D directs entry into LCO 3.0.3 when two or more accumulators are inoperable. This changes the CTS by specifically stating to enter LCO 3.0.3 in this System Specification. This change is acceptable because the action taken when two or more accumulators are inoperable is unchanged. Adding this ACTION is consistent with the ITS convention of directing entry into LCO 3.0.3 when multiple ACTIONS are presented in the ITS, and entry into these multiple ACTIONS could result in a loss of safety function. This change is designated as administrative because it does not result in a technical change to the CTS. A04 CTS 3.5.1.1 ACTION a requires, with one cold leg accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours. CTS ACTION b requires, with one cold leg accumulator inoperable due to boron concentration not within limits, restore boron concentration to within limits within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to 1000 psig or less within the following 6 hours. ITS 3.5.1 ACTION C Required Action C.2 requires, to reduce the RCS pressure to 1000 psig. This changes the CTS by replacing "pressurizer pressure" with "RCS pressure". The purpose of CTS 3.5.1.1 ACTIONS a and b provide compensatory measures for an inoperable accumulator. The change to reduce RCS pressure to 1000 psig better reflects the deviation to the Specification. This change is acceptable because the difference between the pressurizer pressure and RCS pressure is DISCUSSION OF CHANGES ITS 3.5.1, ACCUMULATORS Sequoyah Unit 1 and Unit 2 Page 2 of 3 not significant. This change is designated as administrative because it does not result in a technical change to the CTS. A05 CTS 3.5.1.1.e requires that each cold leg injection accumulator be OPERABLE with power removed from the isolation valve when RCS pressure is above 2000 psig. CTS 4.5.1.1.1.c requires verification that power is removed from the isolation valve operator once per 31 days when RCS pressure is above 2000 psig. ITS SR 3.5.1.5 requires verifying power is removed from each accumulator isolation valve operator when RCS pressure is greater than or equal to 2000 psig. This changes the CTS by requiring power removed from the accumulator isolation valve operators when RCS pressure is at or above 2000 psig. The purpose of CTS 3.5.1.1.e and CTS 4.5.1.1.1.c is to ensure power is removed from each accumulator isolation valve when RCS pressure is above 2000 psig. This change is acceptable because ITS SR 3.5.1.5 requires essentially the same verification of power to be removed from each accumulator isolation valve operator when RCS pressure is greater than or equal to 2000 psig. This change is designated as administrative because it does not result in technical changes to CTS. MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.5.1.1.1.a.1 requires verification of the contained borated water volume and nitrogen cover pressure in each cold leg injection accumulator at least once per 12 hours. CTS 4.5.1.1.1.a.2 requires verification of each cold leg injection accumulator isolation valve is fully open at least once per 12 hours. CTS 4.5.1.1.1.b requires verification of the boron concentration of the cold leg accumulator solution at least once per 31 days. CTS 4.5.1.1.1.c requires verification that power to the isolation valve operator is removed when the RCS pressure is above 2000 psig at least once per 31 days. ITS SR 3.5.1.1, SR 3.5.1.2, SR 3.5.1.3, SR 3.5.1.4, and SR 3.5.1.5 require similar Surveillances and specify the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for these SRs and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information DISCUSSION OF CHANGES ITS 3.5.1, ACCUMULATORS Sequoyah Unit 1 and Unit 2 Page 3 of 3 is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Accumulators 3.5.1 Westinghouse STS 3.5.1-1 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 33.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators

LCO 3.5.1 [Four] ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure > [1000] psig. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator inoperable due to boron concentration not within limits. A.1 Restore boron concentration to within limits. 72 hours B. One accumulator inoperable for reasons other than Condition A. B.1 Restore accumulator to OPERABLE status. 24 hours C. Required Action and associated Completion Time of Condition A or B not met. C.1 Be in MODE 3.

AND C.2 Reduce RCS pressure to [1000] psig. 6 hours

12 hours D. Two or more accumulators inoperable. D.1 Enter LCO 3.0.3. Immediately

3.5.1.1 Applicability ACTION b ACTION a ACTION a ACTION b DOC A03 111 Accumulators 3.5.1Westinghouse STS3.5.1-2Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.1.2 Verify borated water volume in each accumulator is [ 7853 gallons ( )% and 8171 gallons ( )%]. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is [385] psig and [481] psig. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 3.5.1.1.d 4.5.1.1.1.a.1 3.5.1.1.b 4.5.1.1.1.a.1 3.5.1.1.a 4.5.1.1.1.a.2 7615 7960 624 668 21122222 Accumulators 3.5.1Westinghouse STS3.5.1-3Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 3SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify boron concentration in each accumulator is [1900] ppm and [2100] ppm. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] AND --------NOTE-------- Only required to be performed for affected accumulators

Once within 6 hours after each solution volume increase of

[ [  ] gallons, (   )% of indicated level ] that is not the result of addition from the refueling water storage tank SR  3.5.1.5 Verify power is removed from each accumulator isolation valve operator when RCS pressure is 
[2000] psig. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] 3.5.1.1.c 4.5.1.1.1.b 3.5.1.1.e 4.5.1.1.1.c 1% of tank volume 2400 2700 21121223 Accumulators  3.5.1   Westinghouse STS 3.5.1-1 Rev. 4.0   CTS Amendment XXXSEQUOYAH UNIT 2 33.5   EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators

LCO 3.5.1 [Four] ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure > [1000] psig. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator inoperable due to boron concentration not within limits. A.1 Restore boron concentration to within limits. 72 hours B. One accumulator inoperable for reasons other than Condition A. B.1 Restore accumulator to OPERABLE status. 24 hours C. Required Action and associated Completion Time of Condition A or B not met. C.1 Be in MODE 3.

AND C.2 Reduce RCS pressure to [1000] psig. 6 hours

12 hours D. Two or more accumulators inoperable. D.1 Enter LCO 3.0.3. Immediately

3.5.1.1 Applicability ACTION b ACTION a ACTION a ACTION b DOC A03 111 Accumulators 3.5.1Westinghouse STS3.5.1-2Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.1.2 Verify borated water volume in each accumulator is [ 7853 gallons ( )% and 8171 gallons ( )%]. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is [385] psig and [481] psig. [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] 3.5.1.1.d 4.5.1.1.1.a.1 3.5.1.1.b 4.5.1.1.1.a.1 3.5.1.1.a 4.5.1.1.1.a.2 7615 7960 624 668 21122222 Accumulators 3.5.1 Westinghouse STS 3.5.1-3 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 3SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify boron concentration in each accumulator is [1900] ppm and [2100] ppm. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] AND --------NOTE-------- Only required to be performed for affected accumulators

Once within 6 hours after each solution volume increase of

[ [  ] gallons, (   )% of indicated level ] that is not the result of addition from the refueling water storage tank

SR 3.5.1.5 Verify power is removed from each accumulator isolation valve operator when RCS pressure is

[2000] psig.

[ 31 days OR In accordance with the Surveillance Frequency Control Program ] 3.5.1.1.c 4.5.1.1.1.b 3.5.1.1.e 4.5.1.1.1.c 1% of tank volume 2400 2700 21121223 JUSTIFICATION FOR DEVIATIONS ITS 3.5.1, ACCUMULATORS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 2. ISTS SR 3.5.1.1 (ITS SR 3.5.1.1), ISTS SR 3.5.1.2 (ITS SR 3.5.1.2), ISTS SR 3.5.1.3 (ITS SR 3.5.1.3), ISTS SR 3.5.1.4 (ITS SR 3.5.1.4), and ISTS SR 3.5.1.5 (ITS SR 3.5.1.5) provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies for these SRs under the Surveillance Frequency Control Program. 3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Accumulators B 3.5.1 Westinghouse STS B 3.5.1-1 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators

BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA. The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water. The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series. The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. large break large break 11 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-2 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

APPLICABLE The accumulators are assumed OPERABLE in both the large and small SAFETY break LOCA analyses at full power (Ref. 1). These are the Design Basis ANALYSES Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA. The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase. This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following a LOCA:

, safety injection pumps, safety injection and large break  111  each INSERT 1small break and there is a high probability that the criteria are met following a large break LOCA  11 B 3.5.1  Insert Page B 3.5.1-2 INSERT 1   Based on deterministic studies, the worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer.

1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-3 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

APPLICABLE SAFETY ANALYSES (continued)

a. Maximum fuel element cladding temperature is 2200°F,

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation,

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react, and

d. Core is maintained in a coolable geometry. Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. For small breaks, an increase in water volume is a peak clad temperature penalty. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volume from the accumulator to the check valve. The safety analysis assumes values of [6468] gallons and [6879] gallons. To allow for instrument inaccuracy, values of [6520] gallons and [6820] gallons are specified. The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. and the first few seconds of the refill phase2large break 113 7515 8194179607615INSERT 2 realistic large break LOCAtakesbetweenINSERT 3 B 3.5.1 Insert Page B 3.5.1-3 INSERT 2 The large and small break LOCA safety analyses are performed with accumulator volumes that are consistent with the LOCA evaluation models.

INSERT 3 The small break LOCA safety analysis assumes a value from within the range of values used for the large break safety analysis.

11 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-4 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

APPLICABLE SAFETY ANALYSES (continued) The large and small break LOCA analyses are performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 1 and 3). The accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated. For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above [2000] psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist. This LCO is only applicable at pressures > 1000 psig. At pressures 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 2) limit of 2200°F. In MODE 3, with RCS pressure 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators. 2INSERT 4 1 B 3.5.1 Insert Page B 3.5.1-4 INSERT 4 with accumulator pressures that are consistent with the LOCA evaluation models. The realistic large break LOCA safety analysis takes values between 600 psig and 683 psig. To allow for instrument inaccuracy, values of 624 psig and 668 psig are specified. The small break LOCA safety analysis assumes a value from the low end of the range of values taken for the large break safety analysis.

1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-5 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours. In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break for the majority of plants. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours is allowed to return the boron concentration to within limits. B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours. In this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 24 hour Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-15049-A, Rev. 1 (Ref. 4).

C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and RCS pressure reduced to 1000 psig within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. while 1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-6 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

ACTIONS (continued) D.1 If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately. SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully open. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. [ The Frequency of 12 hours is considered reasonable in view of other administrative controls that ensure a mispositioned isolation valve is unlikely. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.1.2 and SR 3.5.1.3 Borated water volume and nitrogen cover pressure are verified for each accumulator. [ The Frequency of 12 hours is sufficient to ensure adequate injection during a LOCA. Because of the static design of the accumulator, a 12 hour Frequency usually allows the operator to identify changes before limits are reached. Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends. OR 454isolation1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-7 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

SURVEILLANCE REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed. [ The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

Sampling the affected accumulator within 6 hours after a 1% volume increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 5). 554 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-8 Rev. 4.0 Revision XXXSEQUOYAH UNIT 1 1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.1.5 Verification that power is removed from each accumulator isolation valve operator when the RCS pressure is [2000] psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. [ Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. REFERENCES 1. FSAR, Chapter [6].

2. 10 CFR 50.46.

3. FSAR, Chapter [15]. 4. WCAP-15049-A, Rev. 1, April 1999. 5. NUREG-1366, February 1990. 45U 21122 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-1 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators

BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA. The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water. The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series. The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. large break large break 11 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-2 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

APPLICABLE The accumulators are assumed OPERABLE in both the large and small SAFETY break LOCA analyses at full power (Ref. 1). These are the Design Basis ANALYSES Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA. The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase. This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following a LOCA:

, safety injection pumps, safety injection and large break  111  each INSERT 1small break and there is a high probability that the criteria are met following a large break LOCA  11 B 3.5.1  Insert Page B 3.5.1-2 INSERT 1   Based on deterministic studies, the worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer.

1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-3 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

APPLICABLE SAFETY ANALYSES (continued)

a. Maximum fuel element cladding temperature is 2200°F,

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation,

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react, and

d. Core is maintained in a coolable geometry. Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. For small breaks, an increase in water volume is a peak clad temperature penalty. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volume from the accumulator to the check valve. The safety analysis assumes values of [6468] gallons and [6879] gallons. To allow for instrument inaccuracy, values of [6520] gallons and [6820] gallons are specified. The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. and the first few seconds of the refill phase2large break 113 7515 8194179607615INSERT 2 realistic large break LOCAtakesbetweenINSERT 3 B 3.5.1 Insert Page B 3.5.1-3 INSERT 2 The large and small break LOCA safety analyses are performed with accumulator volumes that are consistent with the LOCA evaluation models.

INSERT 3 The small break LOCA safety analysis assumes a value from within the range of values used for the large break safety analysis.

11 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-4 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

APPLICABLE SAFETY ANALYSES (continued) The large and small break LOCA analyses are performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 1 and 3). The accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated. For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above [2000] psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist. This LCO is only applicable at pressures > 1000 psig. At pressures 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 2) limit of 2200°F. In MODE 3, with RCS pressure 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators. 2INSERT 4 1 B 3.5.1 Insert Page B 3.5.1-4 INSERT 4 with accumulator pressures that are consistent with the LOCA evaluation models. The realistic large break LOCA safety analysis takes values between 600 psig and 683 psig. To allow for instrument inaccuracy, values of 624 psig and 668 psig are specified. The small break LOCA safety analysis assumes a value from the low end of the range of values taken for the large break safety analysis.

1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-5 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours. In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break for the majority of plants. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours is allowed to return the boron concentration to within limits. B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours. In this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 24 hour Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-15049-A, Rev. 1 (Ref. 4).

C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and RCS pressure reduced to 1000 psig within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. while 1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-6 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

ACTIONS (continued) D.1 If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately. SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully open. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. [ The Frequency of 12 hours is considered reasonable in view of other administrative controls that ensure a mispositioned isolation valve is unlikely. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.1.2 and SR 3.5.1.3 Borated water volume and nitrogen cover pressure are verified for each accumulator. [ The Frequency of 12 hours is sufficient to ensure adequate injection during a LOCA. Because of the static design of the accumulator, a 12 hour Frequency usually allows the operator to identify changes before limits are reached. Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends. OR 454isolation1 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-7 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

SURVEILLANCE REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed. [ The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

Sampling the affected accumulator within 6 hours after a 1% volume increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 5). 554 Accumulators B 3.5.1 Westinghouse STS B 3.5.1-8 Rev. 4.0 Revision XXXSEQUOYAH UNIT 2 1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.1.5 Verification that power is removed from each accumulator isolation valve operator when the RCS pressure is [2000] psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. [ Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. REFERENCES 1. FSAR, Chapter [6].

2. 10 CFR 50.46.

3. FSAR, Chapter [15]. 4. WCAP-15049-A, Rev. 1, April 1999. 5. NUREG-1366, February 1990. 45U 21122 JUSTIFICATION FOR DEVIATIONS ITS 3.5.1 BASES, ACCUMULATORS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. 2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. ISTS 3.5.1 Bases for the Applicable Safety Analysis have been changed to reflect Sequoyah (SQN) specific design. SQN does not give a specific value for instrument uncertainties in large break analysis 4. ISTS SR 3.5.1.1, ISTS SR 3.5.1.2, ISTS SR 3.5.1.3, ISTS SR 3.5.1.4 and ISTS SR 3.5.1.5 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Therefore, the Frequency for ITS SR 3.5.1.1, SR 3.5.1.2, SR 3.5.1.3, SR 3.5.1.4 and SR 3.5.1.5 is "In accordance with the Surveillance Frequency Control Program." 5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.5.1, ACCUMULATORS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 2 ITS 3.5.2, ECCS - OPERATING Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.5.2EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.2 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.2 Two ECCS trains shall be OPERABLE. ---------------------------------------------------------NOTES-------------------------------------------------------------- 1.In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolationvalves for up to 2 hours to perform pressure isolation valve testing per SR 4.4.6.3. 2.In MODE 3, ECCS pumps may be made incapable of injecting to support transition into or fromthe APPLICABILITY of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)System," for up to 4 hours or until the temperature of all RCS cold legs exceeds LTOP armingtemperature (350°F) specified in the PTLR plus 25°F, whichever comes first.--------------------------------------------------------------------------------------------------------------------------------- APPLICABILITY: MODES 1, 2 and 3. ACTION: a.With one or more trains inoperable and with at least 100% of the ECCS flow equivalent to asingle OPERABLE ECCS train available, restore the inoperable train(s) to OPERABLE statuswithin 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOTSHUTDOWN within the following 6 hours.b.With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available,immediately enter LCO 3.0.3.SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS train shall be demonstrated OPERABLE: a.At least once per 12 hours by verifying that the following valves are in the indicated positionswith power to the valve operators removed: January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-3 Amendment No. 28, 36, 86, 140, 276, 299, 326 Page 1 of 4 LCO 3.5.2 LCO 3.5.2 Note 1 LCO 3.5.2 Note 2 Applicability ACTION A ACTION B ACTION C LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.1 M01LA01L01ECCS A01ITS ITS 3.5.2EMERGENCY CORE COOLING SYSTEMS (ECCS) SURVEILLANCE REQUIREMENTS (Continued) Valve Number Valve Function Valve Position a.FCV-63-1RHR Suction from RWST open b.FCV-63-22SIS Discharge to Common Piping open b.At least once per 31 days by:1.Verify ECCS piping is full of water by venting the ECCS pump casings andaccessible piping high points, and2.Verify each ECCS manual, power operated and automatic valve in the flow paththat is not locked, sealed, or otherwise secured in position, is in the correctposition.c.Deletedd.At least once per 18 months perform a visual inspection of the containment sump and verifythat the suction inlets are not restricted by debris and that the sump components (strainers, screens, etc.) show no evidence of structural distress or corrosion. e.At least once per 18 months, by:1.Verifying that each automatic valve in the flow path that is not locked, sealed orotherwise secured in position, actuates to its correct position on an actual orsimulated actuation signal.2.Verifying that each ECCS pump starts automatically on an actual or simulatedactuation signal.f.By verifying that each ECCS pump's developed head at the test flow point is greater thanor equal to the required developed head when tested in accordance with the InserviceTesting Program of Specification 4.0.5.g.At least once per 18 months, verify the correct position of each mechanical stop for thefollowing ECCS throttle valves: Charging Pump Injection Safety Injection Cold Safety Injection Hot Throttle Valves Leg Throttle Valves Leg Throttle Valves Valve Number Valve Number Valve Number 1.63 - 5821.63 - 5501.63-5422.63 - 5832.63 - 5522.63-5443.63 - 5843.63 - 5543.63-5464.63 - 5854.63 - 5564.63-548January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-4 Amendment Nos. 36, 92, 139, 140, 276, 299, 326 Page 2 of 4 SR 3.5.2.1 LA02In accordance with the Surveillance Frequency Control Program LA02In accordance with the Surveillance Frequency Control Program LA02In accordance with the Surveillance Frequency Control Program LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.3 SR 3.5.2.2 SR 3.5.2.8 SR 3.5.2.5 SR 3.5.2.6 SR 3.5.2.5 SR 3.5.2.6 SR 3.5.2.4 SR 3.5.2.7 LA03 A01ITS ITS 3.5.2EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.2 Two ECCS trains shall be OPERABLE. ---------------------------------------------------------NOTES------------------------------------------------------------- 1.In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing theisolation valves for up to 2 hours to perform pressure isolation valve testing per SR 4.4.6.3. 2.In MODE 3, ECCS pumps may be made incapable of injecting to support transition into orfrom the APPLICABILITY of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)System," for up to 4 hours or until the temperature of all RCS cold legs exceeds LTOP armingtemperature (350°F) specified in the PTLR plus 25°F, whichever comes first.-------------------------------------------------------------------------------------------------------------------------------- APPLICABILITY: MODES 1, 2 and 3. ACTION: a.With one or more trains inoperable and with at least 100% of the ECCS flow equivalent to asingle OPERABLE ECCS train available, restore the inoperable train(s) to OPERABLE statuswithin 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOTSHUTDOWN within the following 6 hours.b.With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS trainavailable, immediately enter LCO 3.0.3.SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS train shall be demonstrated OPERABLE: a.At least once per 12 hours by verifying that the following valves are in the indicated positionswith power to the valve operators removed: Valve Number Valve FunctionValve Position a.FCV-63-1RHR Suction from RWST open b.FCV-63-22SIS Discharge to Common Piping open January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-3 Amendment No. 17, 28, 82, 95, 128, 131, 203, 267, 288, 319 LCO 3.5.2 LCO 3.5.2 Note 1 LCO 3.5.2 Note 2 Applicability ACTION A ACTION B ACTION C SR 3.5.2.1 LA02In accordance with the Surveillance Frequency Control Program Page 3 of 4 M01LA01L01ECCS A01ITS ITS 3.5.2EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days by: 1. Verify ECCS piping is full of water by venting the ECCS pump casings and accessible piping high points, and 2. Verify each ECCS manual, power operated and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position. c. Deleted d. At least once per 18 months perform a visual inspection of the containment sump and verify that the suction inlets are not restricted by debris and that the sump components (strainers, screens, etc.) show no evidence of structural distress or corrosion. e. At least once per 18 months, by: 1. Verifying that each automatic valve in the flow path that is not locked, sealed or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal. 2. Verifying that each ECCS pump starts automatically on an actual or simulated actuation signal. f. By verifying that each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head when tested in accordance with the Inservice Testing Program of Specification 4.0.5. g. At least once per 18 months, verify the correct position of each mechanical stop for the following ECCS throttle valves: Charging Pump Injection Safety Injection Cold Safety Injection Hot Throttle Valves Leg Throttle Valves Leg Throttle Valves Valve Number Valve Number Valve Number

1. 63 - 582 1. 63 - 550 1. 63-542 2. 63 - 583 2. 63 - 552 2. 63-544

3. 63 - 584 3. 63 - 554 3. 63-546 4. 63 - 585 4. 63 - 556 4. 63-548

January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-4 Amendment No. 82, 128, 131, 319 LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.3 SR 3.5.2.2 LA03LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.8 LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.5 SR 3.5.2.5 SR 3.5.2.6 SR 3.5.2.6 SR 3.5.2.4 LA02In accordance with the Surveillance Frequency Control Program SR 3.5.2.7 Page 4 of 4 DISCUSSION OF CHANGES ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and Unit 2 Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.5.2 ACTION a entry condition is with one or more trains inoperable and with at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available. ITS 3.5.2 ACTION A entry condition is for one or more trains inoperable. This changes the CTS by deleting the requirement that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available to enter the required action for one or more inoperable ECCS train(s). The purpose of CTS 3.5.2 ACTION a is to limit the time that the plant can continue to operate with one or more ECCS trains inoperable. Stating the entry condition for CTS 3.5.2 ACTION a as a compound condition could cause entry, exiting, and reentry into CTS 3.5.2 ACTION a based on whether the ECCS system has 100% flow capability available with one or more trains inoperable. This compound condition could incorrectly result in an inoperability time beyond 72 hours, without having returned a train to OPERABLE status. CTS 3.5.2 ACTION b Required Actions with less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available are retained in ITS LCO 3.5.2 ACTION C. By splitting CTS 3.5.2 ACTION a entry condition, where ITS LCO 3.5.2 ACTION A is entered when one or more ECCS trains are inoperable and ITS LCO 3.5.2 ACTION C is entered when less than 100% ECCS flow is available, ITS 3.5.2 ACTION A remains applicable regardless of overall remaining ECCS flow availability, so that the completion time clock is not reset in the event 100% flow is restored. This change is acceptable because ITS LCO 3.5.2 ACTION A and ACTION C provide adequate compensatory measures to take with one or more ECCS trains inoperable and address the condition of whether the ECCS system has at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available or not. This change is designated as more restrictive, because it limits the time the plant is allowed to operate with one or more ECCS train inoperable to 72 hours.

RELOCATED SPECIFICATIONS None DISCUSSION OF CHANGES ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and Unit 2 Page 2 of 4 REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS LCO 3.5.2 Note 2 allows an option in MODE 3 for the ECCS pumps to be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for up to 4 hours or until the temperature of all RCS cold legs exceeds LTOP arming temperature (350°) specified in the PTLR plus 25°F. ITS LCO 3.5.2 Note 2 provides the same allowance but does not explicitly include the LTOP arming temperature of 350°F. This changes the CTS by moving the specific value of the LTOP arming temperature (350°) from the CTS to the PTLR. The removal of this detail related to system design from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. Also this change is acceptable, because the removed information is adequately controlled in the PTLR. Changes to the PTLR are controlled in Chapter 5 of the Technical Specifications. This change is designated as a less restrictive removal of detail change because a procedural detail for meeting the TS requirements is being removed from the ITS. LA02 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.5.2.a requires verification that the listed valves are in the indicated positions with power to the valve operators removed once per 12 hours. CTS 4.5.2.b.1 requires verification that the ECCS piping is full of water once per 31 days. CTS 4.5.2.b.2 requires verification that each ECCS manual, power operated and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position once per 31 days. CTS 4.5.2.d requires performance of a visual inspection of the containment sump, suction lines, and sump components once per 18 months. CTS 4.5.2.e.1 requires verification that each automatic valve in the flow path that is not locked, sealed or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal once per 18 months. CTS 4.5.2.e.2 requires verification that the ECCS pump starts automatically on an actual or simulated actuation signal once per 18 months. CTS 4.5.2.g requires verification that each mechanical stop of the ECCS throttle valves is in the correct position once per 18 months. ITS SR 3.5.2.1, ITS SR 3.5.2.2, ITS SR 3.5.2.3, ITS SR 3.5.2.5, ITS SR 3.5.2.6, ITS SR 3.5.2.7 and ITS SR 3.5.2.8 require similar Surveillances and specify the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for these SRs and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the DISCUSSION OF CHANGES ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and Unit 2 Page 3 of 4 control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA03 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 4.5.2.b.1 requires verifying ECCS piping is full of water by venting the ECCS pump casing and accessible piping high points. ITS SR 3.5.2.3 requires verifying ECCS piping is full of water. This changes the CTS by moving the details of how to vent the ECCS piping "by venting the ECCS pump casings and accessible piping high points" from the CTS to the Bases. The removal of these details related to system design from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. Also, this change is acceptable, because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to the Bases to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS LCO 3.5.2 requires, two ECCS trains to be OPERABLE. CTS LCO 3.5.2 Note 1 states, in MODE 3, both safety injection SI pump flow paths may be isolated by closing the isolation valves for up to 2 hours to perform pressure isolation valve testing per SR 4.4.6.3. ITS LCO 3.5.2 requires two ECCS trains to be OPERABLE. ITS LCO 3.5.2 Note 1 states In MODE 3, both ECCS pump flow paths may be isolated for 2 hours to perform pressure isolation valve (PIV) testing per SR 3.4.14.1. This changes the CTS by allowing the RHR pump flow paths to be isolated in addition to the SI pump flow paths for Surveillance testing of the pressure isolation valves. The purpose of CTS LCO 3.5.2 is to ensure that two ECCS trains are OPERABLE in MODES 1, 2, and 3. The purpose of ITS SR 3.4.14.1 is to prevent overpressure failure of the low pressure portions of connecting systems. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, or an unanalyzed accident that could degrade the ability for low pressure injection. PIV testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. Thus ITS SR 3.4.14.1 supports ITS LCO 3.5.2 to ensure that two ECCS trains are OPERABLE. Surveillance testing of the pressure isolation valves requires the SI Pump and RHR Pump flow paths to be isolated. CTS LCO 3.5.2 Note 1 allows DISCUSSION OF CHANGES ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and Unit 2 Page 4 of 4 both SI pump flow paths to be isolated for two hours provided that the flow paths are readily restorable from the control room. In addition to isolating the SI pump flow paths, ITS LCO 3.5.2 Note 1 will allow both RHR pump flow paths to be isolated for two hours allowing for the required testing of the PIVs. This change permits the isolation of the ECCS (SI pump and RHR pump) flow paths provided that the flow paths are readily restorable from the control room. The acceptability of this testing allowance is based on the operability of the centrifugal charging system and the cold leg injection accumulators and the low probability of an accident occurring during the isolation time to support PIV testing. This change is acceptable because the LCO requirements continue to ensure that the system is maintained consistent with the safety analysis and licensing basis. This change is designated as less restrictive because the RHR pump flow path may be isolated to support testing of the PIVs. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) ECCS - Operating 3.5.2Westinghouse STS3.5.2-1Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 43.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE. --------------------------------------------NOTES------------------------------------------- [ 1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours to perform pressure isolation valve testing per SR 3.4.14.1. 2.In MODE 3, ECCS pumps may be made incapable of injecting tosupport transition into or from the Applicability of LCO 3.4.12, "LowTemperature Overpressure Protection (LTOP) System," for up to4 hours or until the temperature of all RCS cold legs exceeds[375°F] [Low Temperature Overpressure Protection (LTOP) armingtemperature specified in the PTLR plus [25]°F], whichever comesfirst. ]-------------------------------------------------------------------------------------------------- APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable. A.1 Restore train(s) to OPERABLE status. 72 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 4. 6 hours 12 hours C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available. C.1 Enter LCO 3.0.3. Immediately 3.5.2 3.5.2 Note 1 3.5.2 Note 2 Applicability ACTION a ACTION a ACTION b 11DOC M01 ECCS4 ECCS - Operating 3.5.2Westinghouse STS3.5.2-2Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 [ Verify the following valves are in the listed position with power to the valve operator removed. NumberPositionFunction [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.3 [ Verify ECCS piping is full of water. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] ] SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head. In accordance with the Inservice Testing Program 4.5.2.a 4.5.2.b.2 4.5.2.b.1 4.5.2.f 222222INSERT 1 1 3.5.2 Insert Page 3.5.2-2 CTS INSERT 1 FCV-63-1 Open RHR Suction from RWST FCV-63-22 Open SIS Discharge to Common Piping 14.5.2.a ECCS - Operating 3.5.2Westinghouse STS3.5.2-3Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 4SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.7 [ Verify, for each ECCS throttle valve listed below, each position stop is in the correct position. Valve Number [ ] [ ] [ ] [ [18] months OR In accordance with the Surveillance Frequency Control Program ] ] 4.5.2.e.1 4.5.2.e.2 4.5.2.g 222222INSERT 2 1mechanical1 3.5.2 Insert Page 3.5.2-3 CTS INSERT 2 Charging Pump Injection Throttle Valves Safety Injection Cold Leg Throttle Valves Safety Injection Hot Leg Throttle Valves 63-58263-550 63-542 63-58363-552 63-544 63-58463-554 63-546 63-58563-556 63-548 14.5.2.g ECCS - Operating 3.5.2Westinghouse STS3.5.2-4Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 4SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.8 Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] 4.5.2.d 22 ECCS - Operating 3.5.2Westinghouse STS3.5.2-1Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 43.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE. --------------------------------------------NOTES------------------------------------------- [ 1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours to perform pressure isolation valve testing per SR 3.4.14.1. 2.In MODE 3, ECCS pumps may be made incapable of injecting tosupport transition into or from the Applicability of LCO 3.4.12, "LowTemperature Overpressure Protection (LTOP) System," for up to4 hours or until the temperature of all RCS cold legs exceeds[375°F] [Low Temperature Overpressure Protection (LTOP) armingtemperature specified in the PTLR plus [25]°F], whichever comesfirst. ]-------------------------------------------------------------------------------------------------- APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable. A.1 Restore train(s) to OPERABLE status. 72 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3. AND B.2 Be in MODE 4. 6 hours 12 hours C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available. C.1 Enter LCO 3.0.3. Immediately 3.5.2 3.5.2 Note 1 3.5.2 Note 2 Applicability ACTION a ACTION a ACTION b 11DOC M01 ECCS4 ECCS - Operating 3.5.2Westinghouse STS3.5.2-2Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 [ Verify the following valves are in the listed position with power to the valve operator removed. NumberPositionFunction [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ 12 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.3 [ Verify ECCS piping is full of water. [ 31 days OR In accordance with the Surveillance Frequency Control Program ] ] SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head. In accordance with the Inservice Testing Program 4.5.2.a 4.5.2.b.2 4.5.2.b.1 4.5.2.f 222222INSERT 1 1 3.5.2 Insert Page 3.5.2-2 CTS INSERT 1 FCV-63-1 Open RHR Suction from RWST FCV-63-22 Open SIS Discharge to Common Piping 14.5.2.a ECCS - Operating 3.5.2Westinghouse STS3.5.2-3Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 4SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.2.7 [ Verify, for each ECCS throttle valve listed below, each position stop is in the correct position. Valve Number [ ] [ ] [ ] [ [18] months OR In accordance with the Surveillance Frequency Control Program ] ] 4.5.2.e.1 4.5.2.e.2 4.5.2.g 222222INSERT 2 1mechanical1 3.5.2 Insert Page 3.5.2-3 CTS INSERT 2 Charging Pump Injection Throttle Valves Safety Injection Cold Leg Throttle Valves Safety Injection Hot Leg Throttle Valves 63-58263-550 63-542 63-58363-552 63-544 63-58463-554 63-546 63-58563-556 63-548 14.5.2.g ECCS - Operating 3.5.2Westinghouse STS3.5.2-4Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 4SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.8 Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion. [ [18] months OR In accordance with the Surveillance Frequency Control Program ] 4.5.2.d 22 JUSTIFICATION FOR DEVIATIONS ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.The ISTS contains bracketed information and/or values that are generic to allWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is inserted to reflect the current licensing basis.2.ISTS SR 3.5.2.1 (ITS SR 3.5.2.1), ISTS SR 3.5.2.2 (ITS SR 3.5.2.2),ISTS SR 3.5.2.3 (ITS SR 3.5.2.3), ISTS SR 3.5.2.5 (ITS SR 3.5.2.5),ISTS SR 3.5.2.6 (ITS SR 3.5.2.6), ISTS SR 3.5.2.7 (ITS SR 3.5.2.7) andISTS SR 3.5.2.8 (ITS SR 3.5.2.8) provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies for these SRs under the Surveillance Frequency Control Program.3.Editorial correction made for clarity.4.Changes made to reflect changes made to the Specification. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-1 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS - Operating

BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents: a. Loss of coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system,

b. Rod ejection accident, c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater, and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold leg recirculation, and hot leg recirculation. In the injection phase, water is taken from the refueling water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation. After approximately 24 hours, the ECCS flow is shifted to the hot leg recirculation phase to provide a backflush, which would reduce the boiling in the top of the core and any resulting boron precipitation. The ECCS consists of three separate subsystems: centrifugal charging (high head), safety injection (SI) (intermediate head), and residual heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains. The ECCS accumulators and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by this LCO. 5.5 1INSERT 1 1 Insert Page B 3.5.2-1 INSERT 1 There are two modes of ECCS operation, injection and recirculation. The injection mode consists of the injection phase and the recirculation mode consists of the cold leg recirculation phase and hot leg recirculation phase. 1 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES BACKGROUND (continued) The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the RWST can be injected into the RCS following the accidents described in this LCO. The major components of each subsystem are the centrifugal charging pumps, the RHR pumps, heat exchangers, and the SI pumps. Each of the three subsystems consists of two 100% capacity trains that are interconnected and redundant such that either train is capable of supplying 100% of the flow required to mitigate the accident consequences. This interconnecting and redundant subsystem design provides the operators with the ability to utilize components from opposite trains to achieve the required 100% flow to the core. During the injection phase of LOCA recovery, a suction header supplies water from the RWST to the ECCS pumps. Separate piping supplies each subsystem and each train within the subsystem. The discharge from the centrifugal charging pumps combines prior to entering the boron injection tank (BIT) (if the plant utilizes a BIT) and then divides again into four supply lines, each of which feeds the injection line to one RCS cold leg. The discharge from the SI and RHR pumps divides and feeds an injection line to each of the RCS cold legs. Control valves are set to balance the flow to the RCS. This balance ensures sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs. For LOCAs that are too small to depressurize the RCS below the shutoff head of the SI pumps, the centrifugal charging pumps supply water until the RCS pressure decreases below the SI pump shutoff head. During this period, the steam generators are used to provide part of the core cooling function. During the recirculation phase of LOCA recovery, RHR pump suction is transferred to the containment sump. The RHR pumps then supply the other ECCS pumps. Initially, recirculation is through the same paths as the injection phase. Subsequently, recirculation alternates injection between the hot and cold legs.

The centrifugal charging subsystem of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the negative moderator temperature coefficient is highly negative, such as at the end of each cycle. centrifugal charging pump injection tank (CCPIT) 1RHR Throttle 1and SIs16 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-3 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES BACKGROUND (continued) During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.

The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet GDC 35 (Ref. 1). APPLICABLE The LCO helps to ensure that the following acceptance criteria for the SAFETY ECCS, established by 10 CFR 50.46 (Ref. 2), will be met following a ANALYSES LOCA: a. Maximum fuel element cladding temperature is 2200°F,

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation,

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react, d. Core is maintained in a coolable geometry, and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment temperature limits are met. 1 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-4 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

APPLICABLE SAFETY ANALYSES (continued) Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Refs. 3 and 4). This event establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The centrifugal charging pumps and SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The SGTR and MSLB events also credit the centrifugal charging pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions: a. A large break LOCA event, with loss of offsite power and a single failure disabling one RHR pump (both EDG trains are assumed to operate due to requirements for modeling full active containment heat removal system operation) and

b. A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.

The effects on containment mass and energy releases are accounted for in appropriate analyses (Refs. 3 and 4). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core being uncovered following a large LOCA. It also ensures that the centrifugal charging and SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the centrifugal charging pump delivers sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents. requiredECCS train ECCS 1111(both containment spray trains are assumed to operate conservatively reducing containment pressure and increasing break flow) break ECCS - Operating B 3.5.2 Westinghouse STSB 3.5.2-5Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES LCO (continued) In MODES 1, 2, and 3, an ECCS train consists of a centrifugal charging subsystem, an SI subsystem, and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and automatically transferring suction to the containment sump. During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains. As indicated in Note 1, the SI flow paths may be isolated for 2 hours in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room. As indicated in Note 2, operation in MODE 3 with ECCS trains made incapable of injecting in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," is necessary for plants with an LTOP arming temperature at or near the MODE 3 boundary temperature of 350°F. LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the LTOP arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the LTOP Applicability, and provide time to restore the inoperable pumps to OPERABLE status on exiting the LTOP Applicability. APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA. MODE 2 and MODE 3 requirements are bounded by the MODE 1 analysis. ECCSECCS11RHR1ECCS (SI Pump and RHR Pump) 7 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-6 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES APPLICABILITY (continued) This LCO is only applicable in MODE 3 and above. Below MODE 3, the SI signal setpoint is manually bypassed by operator control, and system functional requirements are relaxed as described in LCO 3.5.3, "ECCS - Shutdown."

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A.1 With one or more trains inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours. The 72 hour Completion Time is based on an NRC reliability evaluation (Ref. 5) and is a reasonable time for repair of many ECCS components. An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable. An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 5) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours. ECCSECCS112 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-7 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

ACTIONS (continued) Reference 6 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered. B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 Condition A is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available. With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the facility is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately. SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power or by key locking the control in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned. These valves are of the type, described in Reference 6, that can disable the function of both ECCS trains and invalidate the accident analyses. [ A 12 hour Frequency is considered reasonable in view of other administrative controls that will ensure a mispositioned valve is unlikely. 31 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-8 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. [ The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

3443 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-9 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.2.3 With the exception of the operating centrifugal charging pump, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases. Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an SI signal or during shutdown cooling. [ The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the procedural controls governing system operation. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the Inservice Testing Program of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements. Venting of the ECCS piping is accomplished by venting the pump casings and accessible high point vents. 431 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-10 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. [ The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.2.7 Realignment of valves in the flow path on an SI signal is necessary for proper ECCS performance. These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance is not required for plants with flow limiting orifices. [ The 18 month Frequency is based on the same reasons as those stated in SR 3.5.2.5 and SR 3.5.2.6. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 4336 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-11 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued)

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.2.8 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. [ The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to have access to the location, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.

2. 10 CFR 50.46.

3. FSAR, Section [ ].

4. FSAR, Chapter [15], "Accident Analysis." 5. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

6. IE Information Notice No. 87-01. U 6.343451 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-1 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS - Operating

BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents: a. Loss of coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system,

b. Rod ejection accident, c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater, and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold leg recirculation, and hot leg recirculation. In the injection phase, water is taken from the refueling water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation. After approximately 24 hours, the ECCS flow is shifted to the hot leg recirculation phase to provide a backflush, which would reduce the boiling in the top of the core and any resulting boron precipitation. The ECCS consists of three separate subsystems: centrifugal charging (high head), safety injection (SI) (intermediate head), and residual heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains. The ECCS accumulators and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by this LCO. 5.5 1INSERT 1 1 Insert Page B 3.5.2-1 INSERT 1 There are two modes of ECCS operation, injection and recirculation. The injection mode consists of the injection phase and the recirculation mode consists of the cold leg recirculation phase and hot leg recirculation phase. 1 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES BACKGROUND (continued) The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the RWST can be injected into the RCS following the accidents described in this LCO. The major components of each subsystem are the centrifugal charging pumps, the RHR pumps, heat exchangers, and the SI pumps. Each of the three subsystems consists of two 100% capacity trains that are interconnected and redundant such that either train is capable of supplying 100% of the flow required to mitigate the accident consequences. This interconnecting and redundant subsystem design provides the operators with the ability to utilize components from opposite trains to achieve the required 100% flow to the core. During the injection phase of LOCA recovery, a suction header supplies water from the RWST to the ECCS pumps. Separate piping supplies each subsystem and each train within the subsystem. The discharge from the centrifugal charging pumps combines prior to entering the boron injection tank (BIT) (if the plant utilizes a BIT) and then divides again into four supply lines, each of which feeds the injection line to one RCS cold leg. The discharge from the SI and RHR pumps divides and feeds an injection line to each of the RCS cold legs. Control valves are set to balance the flow to the RCS. This balance ensures sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs. For LOCAs that are too small to depressurize the RCS below the shutoff head of the SI pumps, the centrifugal charging pumps supply water until the RCS pressure decreases below the SI pump shutoff head. During this period, the steam generators are used to provide part of the core cooling function. During the recirculation phase of LOCA recovery, RHR pump suction is transferred to the containment sump. The RHR pumps then supply the other ECCS pumps. Initially, recirculation is through the same paths as the injection phase. Subsequently, recirculation alternates injection between the hot and cold legs.

The centrifugal charging subsystem of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the negative moderator temperature coefficient is highly negative, such as at the end of each cycle. centrifugal charging pump injection tank (CCPIT) 1RHR Throttle 1and SIs16 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-3 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES BACKGROUND (continued) During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.

The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet GDC 35 (Ref. 1). APPLICABLE The LCO helps to ensure that the following acceptance criteria for the SAFETY ECCS, established by 10 CFR 50.46 (Ref. 2), will be met following a ANALYSES LOCA: a. Maximum fuel element cladding temperature is 2200°F,

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation,

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react, d. Core is maintained in a coolable geometry, and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment temperature limits are met. 1 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-4 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

APPLICABLE SAFETY ANALYSES (continued) Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Refs. 3 and 4). This event establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The centrifugal charging pumps and SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The SGTR and MSLB events also credit the centrifugal charging pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions: a. A large break LOCA event, with loss of offsite power and a single failure disabling one RHR pump (both EDG trains are assumed to operate due to requirements for modeling full active containment heat removal system operation) and

b. A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.

The effects on containment mass and energy releases are accounted for in appropriate analyses (Refs. 3 and 4). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core being uncovered following a large LOCA. It also ensures that the centrifugal charging and SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the centrifugal charging pump delivers sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents. requiredECCS train ECCS 1111(both containment spray trains are assumed to operate conservatively reducing containment pressure and increasing break flow) break ECCS - Operating B 3.5.2 Westinghouse STSB 3.5.2-5Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES LCO (continued) In MODES 1, 2, and 3, an ECCS train consists of a centrifugal charging subsystem, an SI subsystem, and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and automatically transferring suction to the containment sump. During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains. As indicated in Note 1, the SI flow paths may be isolated for 2 hours in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room. As indicated in Note 2, operation in MODE 3 with ECCS trains made incapable of injecting in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," is necessary for plants with an LTOP arming temperature at or near the MODE 3 boundary temperature of 350°F. LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the LTOP arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the LTOP Applicability, and provide time to restore the inoperable pumps to OPERABLE status on exiting the LTOP Applicability. APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA. MODE 2 and MODE 3 requirements are bounded by the MODE 1 analysis. ECCSECCS11RHR1ECCS (SI Pump and RHR Pump) 7 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-6 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES APPLICABILITY (continued) This LCO is only applicable in MODE 3 and above. Below MODE 3, the SI signal setpoint is manually bypassed by operator control, and system functional requirements are relaxed as described in LCO 3.5.3, "ECCS - Shutdown."

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A.1 With one or more trains inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours. The 72 hour Completion Time is based on an NRC reliability evaluation (Ref. 5) and is a reasonable time for repair of many ECCS components. An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable. An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 5) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours. ECCSECCS112 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-7 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

ACTIONS (continued) Reference 6 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered. B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 Condition A is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available. With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the facility is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately. SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power or by key locking the control in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned. These valves are of the type, described in Reference 6, that can disable the function of both ECCS trains and invalidate the accident analyses. [ A 12 hour Frequency is considered reasonable in view of other administrative controls that will ensure a mispositioned valve is unlikely. 31 ECCS - Operating B 3.5.2 Westinghouse STS B 3.5.2-8 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. [ The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

3443 ECCS - Operating B 3.5.2 Westinghouse STSB 3.5.2-9Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.5.2.3 With the exception of the operating centrifugal charging pump, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases. Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an SI signal or during shutdown cooling. [ The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the procedural controls governing system operation. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the Inservice Testing Program of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements. Venting of the ECCS piping is accomplished by venting the pump casings and accessible high point vents. 431 ECCS - Operating B 3.5.2 Westinghouse STSB 3.5.2-10Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. [ The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.2.7 Realignment of valves in the flow path on an SI signal is necessary for proper ECCS performance. These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance is not required for plants with flow limiting orifices. [ The 18 month Frequency is based on the same reasons as those stated in SR 3.5.2.5 and SR 3.5.2.6. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 4336 ECCS - Operating B 3.5.2 Westinghouse STSB 3.5.2-11Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES SURVEILLANCE REQUIREMENTS (continued) -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.2.8 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. [ The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to have access to the location, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] REFERENCES 1.10 CFR 50, Appendix A, GDC 35.2.10 CFR 50.46.

3. FSAR, Section[ ].4. FSAR, Chapter [15], "Accident Analysis."5.NRC Memorandum to V. Stello, Jr., from R.L. Baer, "RecommendedInterim Revisions to LCOs for ECCS Components,"December 1, 1975.6.IE Information Notice No. 87-01.U 6.343451 JUSTIFICATION FOR DEVIATIONS ITS 3.5.2 BASES, ECCS OPERATING Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS Bases thatreflect the plant-specific nomenclature, number, reference, system description,analysis, or licensing basis description.2.The listed LCOs concern the shutdown cooling function of the RHR System, not theECCS function. The Applicability Section has adequately described why ECCS is not needed in MODES 5 and 6, and it is not necessary to describe why normal shutdown cooling is required. Therefore, this inappropriate information has been deleted.3.ISTS SR 3.5.2.1, ISTS SR 3.5.2.2, ISTS SR 3.5.2.3, ISTS SR 3.5.2.5, ISTS SR3.5.2.6, ISTS SR 3.5.2.7 and ISTS SR 3.5.2.8 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program.

Therefore, the Frequency for ITS SR 3.5.2.1, SR 3.5.2.2, SR 3.5.2.3, SR 3.5.2.5, SR3.5.2.6, SR 3.5.2.7 and SR 3.5.2.8 is "In accordance with the SurveillanceFrequency Control Program."4.The Reviewer's Note has been deleted. This information is for the NRC reviewer tobe keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.5.The ISTS contains bracketed information and/or values that are generic toWestinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.6.Editorial changes made to enhance clarity/consistency.7.Changes are made to reflect changes made to the Specification. Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.5.2, ECCS - OPERATING Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 3 ITS 3.5.3, ECCS - SHUTDOWN Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.5.3EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.3 ECCS -SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.3 One ECCS train shall be OPERABLE. -------------------------------------------------------------NOTE-------------------------------------------------------------------- An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation. ----------------------------------------------------------------------------------------------------------------------------------------- APPLICABILITY: MODE 4. ACTION: -------------------------------------------------------------NOTE-------------------------------------------------------------------- 1. LCO 3.0.4b is not applicable to ECCS centrifugal charging subsystem. 2. The required ECCS residual heat removal (RHR) subsystem may be inoperable for up to 1 hour for surveillance testing of valves provided that alternate heat removal methods are available via the steam generators to maintain reactor coolant system Tavg less than 350°F and provided that the required subsystem is capable of being manually realigned to the ECCS mode of operation. ----------------------------------------------------------------------------------------------------------------------------------------- a. With the required ECCS residual heat removal (RHR) subsystem inoperable, immediately initiate action to restore required ECCS RHR subsystem to OPERABLE status. b. With the required ECCS centrifugal charging subsystem inoperable, within one hour, restore required ECCS centrifugal charging subsystem to OPERABLE status, or be in COLD SHUTDOWN within 24 hours. SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS train shall be demonstrated OPERABLE per the following applicable Surveillance Requirements of 4.5.2: SR 4.5.2.b.1 SR 4.5.2.d SR 4.5.2.f SR 4.5.2.g

January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-5 Amendment No. 36, 140, 276, 301, 326 Page 1 of 4 Applicability LCO 3.5.3 LCO 3.5.3 Note 1 LCO 3.5.3 ACTIONS Note ACTION A ACTION B ACTION C SR 3.5.3.1 LCO 3.5.3 Note 2 A02 A01ITS ITS 3.5.3 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.4 DELETED LIMITING CONDITION FOR OPERATION This Specification is deleted.

January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-6 Amendment No. 140, 326 Page 2 of 4 A01ITS ITS 3.5.3EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS -SHUTDOWN

LIMITING CONDITION FOR OPERATION

3.5.3 One ECCS train shall be OPERABLE. ---------------------------------------------------------NOTE-------------------------------------------------------------------- An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation. ------------------------------------------------------------------------------------------------------------------------------------- APPLICABILITY: MODE 4. ACTION: ---------------------------------------------------------NOTE-------------------------------------------------------------------- 1. LCO 3.0.4b is not applicable to ECCS centrifugal charging subsystem. 2. The required ECCS residual heat removal (RHR) subsystem may be inoperable for up to 1 hour for surveillance testing of valves provided that alternate heat removal methods are available via the steam generators to maintain reactor coolant system Tavg less than 350°F and provided that the required subsystem is capable of being manually realigned to the ECCS mode of operation. ------------------------------------------------------------------------------------------------------------------------------------- a. With the required ECCS residual heat removal (RHR) subsystem inoperable, immediately initiate action to restore required ECCS RHR subsystem to OPERABLE status. b. With the required ECCS centrifugal charging subsystem inoperable, within one hour, restore required ECCS centrifugal charging subsystem to OPERABLE status, or be in COLD SHUTDOWN within 24 hours. SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS train shall be demonstrated OPERABLE per the following applicable Surveillance Requirements of 4.5.2: SR 4.5.2.b.1 SR 4.5.2.d SR 4.5.2.f SR 4.5.2.g

January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-5 Amendment No. 28, 131, 267, 290, 319 LCO 3.5.3 LCO 3.5.3 Note 1 Applicability LCO 3.5.3 ACTIONS Note ACTION A ACTION B ACTION C SR 3.5.3.1 LCO 3.5.3 Note 2 A02Page 3 of 4 A01ITS ITS 3.5.3EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 DELETED

LIMITING CONDITION FOR OPERATION This Specification is deleted.

January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-6 Amendment No. 131, 319 Page 4 of 4 DISCUSSION OF CHANGES ITS 3.5.3, ECCS - SHUTDOWN Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.5.3 ACTION Note 2 states, the required ECCS residual heat removal (RHR) subsystem may be inoperable for up to 1 hour for surveillance testing of valves provided that alternate heat removal methods are available via the steam generators to maintain reactor coolant system Tavg less than 350° F and provided that the required subsystem is capable of being manually realigned to the ECCS mode of operation. ITS LCO 3.5.3 Note 2 states, the required ECCS residual heat removal (RHR) subsystem may be inoperable for up to 1 hour for surveillance testing of valves provided the required subsystem is capable of being manually realigned to the ECCS mode of operation. This changes the CTS by removing the requirement for alternate heat removal methods from the ECCS Specification. ITS LCO 3.4.6 retains the requirements for decay heat removal in MODE 4. This change is acceptable because the CTS requirements have not changed. ITS LCO 3.4.6 retains the requirements for decay heat removal in MODE 4. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES None

LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) ECCS - Shutdown 3.5.3Westinghouse STS 3.5.3-1 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 13.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. ---------------------------------------------NOTE-------------------------------------------- An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation. -------------------------------------------------------------------------------------------------- APPLICABILITY: MODE 4. ACTIONS ------------------------------------------------------------NOTE----------------------------------------------------------- LCO 3.0.4.b is not applicable to ECCS high head subsystem. ------------------------------------------------------------------------------------------------------------------------------- CONDITION REQUIRED ACTION COMPLETION TIME A. [ Required ECCS residual heat removal (RHR) subsystem inoperable. A.1 Initiate action to restore required ECCS RHR subsystem to OPERABLE status. Immediately ] B. Required ECCS [high head subsystem] inoperable. B.1 Restore required ECCS [high head subsystem] to OPERABLE status. 1 hour C. Required Action and associated Completion Time [of Condition B] not met. C.1 Be in MODE 5. 24 hours 3.5.3 3.5.3 Note 1 Applicability ACTION Note centrifugal charging ACTION a ACTION b ACTION b INSERT 1 123331. Residual Heat Removal (RHR)4centrifugal charging centrifugal charging2S4 3.5.3Insert Page 3.5.3-1 CTS INSERT 1 2.The required ECCS residual heat removal (RHR) subsystem may be inoperable for up to1 hour for surveillance testing of valves provided that the required subsystem is capable ofbeing manually realigned to the ECCS mode of operation.23.5.3 Note 2 ECCS - Shutdown 3.5.3Westinghouse STS 3.5.3-2 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 1SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all equipment required to be OPERABLE: [SR 3.5.2.1][SR 3.5.2.7] [SR 3.5.2.3] SR 3.5.2.8 SR 3.5.2.4 In accordance with applicable SRs 4.5.3 4.5.2.b.1 4.5.2.d 4.5.2.f 4.5.2.g 3 ECCS - Shutdown 3.5.3Westinghouse STS 3.5.3-1 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 13.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. ---------------------------------------------NOTE-------------------------------------------- An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation. -------------------------------------------------------------------------------------------------- APPLICABILITY: MODE 4. ACTIONS ------------------------------------------------------------NOTE----------------------------------------------------------- LCO 3.0.4.b is not applicable to ECCS high head subsystem. ------------------------------------------------------------------------------------------------------------------------------- CONDITION REQUIRED ACTION COMPLETION TIME A. [ Required ECCS residual heat removal (RHR) subsystem inoperable. A.1 Initiate action to restore required ECCS RHR subsystem to OPERABLE status. Immediately ] B. Required ECCS [high head subsystem] inoperable. B.1 Restore required ECCS [high head subsystem] to OPERABLE status. 1 hour C. Required Action and associated Completion Time [of Condition B] not met. C.1 Be in MODE 5. 24 hours 3.5.3 3.5.3 Note 1 Applicability ACTION Note centrifugal charging ACTION a ACTION b ACTION b INSERT 1 123331. Residual Heat Removal (RHR)4centrifugal charging centrifugal charging2S4 3.5.3Insert Page 3.5.3-1 CTS INSERT 1 2.The required ECCS residual heat removal (RHR) subsystem may be inoperable for up to1 hour for surveillance testing of valves provided that the required subsystem is capable ofbeing manually realigned to the ECCS mode of operation.23.5.3 Note 2 ECCS - Shutdown 3.5.3Westinghouse STS 3.5.3-2 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 1SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all equipment required to be OPERABLE: [SR 3.5.2.1][SR 3.5.2.7] [SR 3.5.2.3] SR 3.5.2.8 SR 3.5.2.4 In accordance with applicable SRs 4.5.3 4.5.2.b.1 4.5.2.d 4.5.2.f 4.5.2.g 3 JUSTIFICATION FOR DEVIATIONS ITS 3.5.3, ECCS - SHUTDOWN Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Changes are made (additions, deletions, and/or changes) to the ISTS that reflect theplant specific nomenclature, number, reference, system description, analysis, orlicensing basis description2.LCO 3.5.3 Note 2 has been added to be consistent with Sequoyah current licensingbasis. Note 2 allows the required ECCS residual heat removal (RHR) subsystem tobe inoperable for up to 1 hour for surveillance testing of valves provided the requiredsubsystem is capable of being manually realigned to the ECCS mode of operation (CTS 3.5.3 Action Note 2).3.The ISTS contains bracketed information and/or values that are generic to allWestinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.4.Editorial correction made for clarity. Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-1 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS - Shutdown

BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications. In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head).

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2. APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also applies to this SAFETY Bases section. ANALYSES Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA. In MODE 4, an ECCS train consists of a centrifugal charging subsystem and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs. and adequate core cooling is maintained 1RHR1 ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

LCO (continued) This LCO is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation in the RHR mode during MODE 4. APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2. In MODE 4 with RCS temperature below 350°F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements. In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS high head subsystem when entering MODE 4. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS high head subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers. The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous. centrifugal charging 1213two Notes. Note 1INSERT 1 88 Insert Page B 3.5.3-2 INSERT 1 A second Note allows the required ECCS RHR subsystem to be inoperable because of surveillance testing of RCS pressure isolation valve leakage (FCV-74-1 and FCV-74-2). This allows testing while RCS pressure is high enough to obtain valid leakage data and following valve closure for RHR decay heat removal path. The condition requiring manual realignment capability (FCV-74-1 and FCV-74-2 can be opened from the main control room) ensures that in the unlikely event of a DBA during the one hour of surveillance testing, the RHR subsystem can be placed in ECCS recirculation mode when required to mitigate the event. 1 ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-3 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

ACTIONS (continued) With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status. B.1 With no ECCS high head subsystem OPERABLE, due to the inoperability of the centrifugal charging pump or flow path from the RWST, the plant is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour Completion Time to restore at least one ECCS high head subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required. C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators. SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply. REFERENCES The applicable references from Bases 3.5.2 apply. subsystems centrifugal charging centrifugal charging the plant should be placed in MODE 5 415167 ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-1 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS - Shutdown

BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications. In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head).

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2. APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also applies to this SAFETY Bases section. ANALYSES Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA. In MODE 4, an ECCS train consists of a centrifugal charging subsystem and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs. and adequate core cooling is maintained 1RHR1 ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

LCO (continued) This LCO is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation in the RHR mode during MODE 4. APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2. In MODE 4 with RCS temperature below 350°F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements. In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS high head subsystem when entering MODE 4. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS high head subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers. The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous. centrifugal charging 1213two Notes. Note 1INSERT 1 88 Insert Page B 3.5.3-2 INSERT 1 A second Note allows the required ECCS RHR subsystem to be inoperable because of surveillance testing of RCS pressure isolation valve leakage (FCV-74-1 and FCV-74-2). This allows testing while RCS pressure is high enough to obtain valid leakage data and following valve closure for RHR decay heat removal path. The condition requiring manual realignment capability (FCV-74-1 and FCV-74-2 can be opened from the main control room) ensures that in the unlikely event of a DBA during the one hour of surveillance testing, the RHR subsystem can be placed in ECCS recirculation mode when required to mitigate the event. 1 ECCS - Shutdown B 3.5.3 Westinghouse STS B 3.5.3-3 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

ACTIONS (continued) With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status. B.1 With no ECCS high head subsystem OPERABLE, due to the inoperability of the centrifugal charging pump or flow path from the RWST, the plant is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour Completion Time to restore at least one ECCS high head subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required. C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators. SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply. REFERENCES The applicable references from Bases 3.5.2 apply. subsystems centrifugal charging centrifugal charging the plant should be placed in MODE 5 415167 JUSTIFICATION FOR DEVIATIONS ITS 3.5.3 BASES, ECCS - SHUTDOWN Sequoyah Unit 1 and Unit 2 Page 1 of 2 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. 2. The MODE 5 and MODE 6 core cooling LCOs listed are relative to the shutdown cooling function of the RHR System and not the ECCS function. The Applicability Section has adequately described why ECCS is not needed in MODES 5 and 6, and it is not necessary to address the requirements of the shutdown cooling function. Therefore, this inappropriate information has been deleted. 3. The statement in ACTION A.1 Bases concerning how decay heat is removed is not appropriate for this Specification since ITS 3.5.3 is relative to ECCS and not decay heat removal. Normal decay heat removal in MODE 4 is addressed in ITS LCO 3.4.6. In addition, Required Action A.1 of the Specification addresses the requirements to restore the ECCS RHR subsystem for ECCS purposes and not normal decay heat removal. Therefore, the statements discussing decay heat removal have been deleted. 4. ISTS 3.5.3 ACTION A.1 Bases states "With both RHR pumps and heat exchangers inoperable..." ITS 3.5.3 ACTION A.1 Bases states "With both RHR subsystems inoperable..." This changes the ISTS 3.5.3 ACTION A.1 Bases by expanding the reasons that a RHR subsystem may be inoperable beyond a pump and/or heat exchanger being inoperable. This is acceptable, since there may be other reasons that both RHR subsystems are inoperable, and the statement that both RHR subsystems are inoperable is sufficient and is consistent with the actual wording of ITS Required Action A.1. In addition, the required components of an OPERABLE RHR subsystem, including pumps and heat exchangers, are defined in other sections of the ITS 3.5.3 Bases, including the third paragraph of the Background section, and the second paragraph of the LCO section. 5. ISTS 3.5.3 ACTION B.1 Bases states "With no ECCS high head subsystem OPERABLE, due to the inoperability of the centrifugal charging pump or flow path of the RWST..." ITS 3.5.3 ACTION B.1 Bases states "With no ECCS centrifugal charging subsystem OPERABLE..." This changes the ISTS 3.5.3 ACTION B.1 Bases by deleting the statement concerning how a centrifugal charging subsystem is determined to be inoperable. This is acceptable, since there may be other reasons that the ECCS centrifugal charging subsystem is inoperable, and the statement that the ECCS centrifugal charging subsystem is inoperable is sufficient and is consistent with the actual wording of ITS Required Action B.1. In addition, the required components of an OPERABLE centrifugal charging subsystem, including pumps and suction source, are defined in other sections of the ITS 3.5.3 Bases, including the third paragraph of the Background section, and the second paragraph of the LCO section. 6. The statement in ACTION B.1 Bases regarding initiation of actions to place the plant in MODE 5 has been deleted, since the statement is not consistent with the actual wording of ITS Required Action B.1. ITS Required Action B.1 does not address a plant cooldown to MODE 5; it only addresses restoring the inoperable ECCS subsystem to OPERABLE status. ITS Required Action C.1 contains the requirement to place the unit in MODE 5. JUSTIFICATION FOR DEVIATIONS ITS 3.5.3 BASES, ECCS - SHUTDOWN Sequoyah Unit 1 and Unit 2 Page 2 of 2 7. ISTS ACTION C.1 states, in part, that a controlled shutdown should be initiated when the Required Actions of Condition B cannot be completed within the required Completion Time. ITS ACTION C.1, states, in part, that the plant should be placed in MODE 5 if the Required Actions of Condition B cannot be completed within the required Completion Time. This change is acceptable since the statement is consistent with the actual wording of ITS Required Action C.1 and is a more accurate action statement than the ISTS Bases statement that a controlled shutdown should be initiated. 8. Changes have been made to reflect changes made to Specification.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.5.3, ECCS - SHUTDOWN Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 4 ITS 3.5.4, REFUELING WATER STORAGE TANK Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.5.4EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A contained borated water volume of between 370,000 and 375,000 gallons,

b. A boron concentration of between 2500 and 2700 ppm of boron, c. A minimum solution temperature of 60°F, and

d. A maximum solution temperature of 105°F. APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by: 1. Verifying the contained borated water volume in the tank, and 2. Verifying the boron concentration of the water. b. At least once per 24 hours by verifying the RWST temperature.

January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-7 Amendment No. 12, 140, 326 LCO 3.5.4 SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.1 SR 3.5.4.1 Applicability ACTION B ACTION C LA01In accordance with the Surveillance Frequency Control Program SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.1 LA01In accordance with the Surveillance Frequency Control Program Page 1 of 2 SR 3.5.4.2 SR 3.5.4.3 Add proposed Action AL01for reasons other than Condition A LA02A02greater than or equal to A01ITS ITS 3.5.4EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK

LIMITING CONDITION FOR OPERATION

3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A contained borated water volume of between 370,000 and 375,000 gallons,

b. A boron concentration of between 2500 and 2700 ppm of boron,

c. A minimum solution temperature of 60°F, and

d. A maximum solution temperature of 105°F.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REQUIREMENTS

4.5.5 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the contained borated water volume in the tank, and

2. Verifying the boron concentration of the water.

b. At least once per 24 hours by verifying the RWST temperature.

January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-7 Amendment No. 131, 319 LCO 3.5.4 SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.1 SR 3.5.4.1 Applicability ACTION B ACTION C Add proposed Action Afor reasons other than Condition A L01LA01In accordance with the Surveillance Frequency Control Program LA01In accordance with the Surveillance Frequency Control Program SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.2 SR 3.5.4.3 SR 3.5.4.1 Page 2 of 2 LA02A02greater than or equal to DISCUSSION OF CHANGES ITS 3.5.4, REFUELING WATER STORAGE TANK Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 3.5.5 a requires, in part, the refueling water storage tank (RWST) shall be OPERABLE with a contained borated water volume of between 370,000 and 375,000 gallons. ITS SR 3.5.4.2 requires, in part, a similar requirement of the RWST volume of greater than or equal to 370,000 gallons. This changes the CTS by explicitly requiring the RWST minimum volume to be greater than or equal to 370,000 gallons. The discussion for removing the maximum RWST borated volume of 375,000 gallons is contained in DOC LA02. This change is acceptable because no changes are made to the CTS requirements. The change in format from the CTS to the ITS maintains the technical requirements of the minimum required RWST level. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.5.5.a.1 requires verifying the contained borated water volume in the RWST tank once per 7 days. CTS 4.5.5.a.2 requires verifying the boron concentration in the RWST once per 7 days. CTS 4.5.5.b requires the verification of the RWST temperature once per 24 hours. ITS SR 3.5.4.1, SR 3.5.4.2, and SR 3.5.4.3 require similar Surveillances and specify the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequencies for these SRs and associated Bases to the Surveillance Frequency Control Program. DISCUSSION OF CHANGES ITS 3.5.4, REFUELING WATER STORAGE TANK Sequoyah Unit 1 and Unit 2 Page 2 of 3 The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. LA02 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, or ISI Program) CTS 3.5.5 a requires the refueling water storage tank (RWST) to be OPERABLE with a contained borated water volume of between 370,000 and 375,000 gallons. ITS SR 3.5.4.2 requires verification of RWST borated water volume is greater than or equal to 370,000 gallons. This changes the CTS by moving the stated maximum RWST borated water volume of 375,000 to the SQN UFSAR. The removal of these details related to system design from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement to verify the minimum required RWST borated water volume. This change is acceptable because the UFSAR contains the maximum borated water volume for the RWST. The removed requirements will be adequately controlled in the UFSAR as any changes to the UFSAR are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change, because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.5.5 ACTION allows 1 hour to restore an inoperable RWST. ITS 3.5.4 ACTION A allows 8 hours to restore the RWST to OPERABLE status if the inoperability is due to the RWST boron concentration or temperature not within limits. ITS 3.5.4 ACTION B requires the restoration of the RWST to an OPERABLE status within 1 hour for reasons other than Condition A. This changes the CTS by increasing the Completion Time for restoration of an inoperable RWST due to boron concentration or temperature not within limits from 1 hour to 8 hours. The purpose of CTS 3.5.5 Action is to require rapid correction of conditions that affect both trains of ECCS. This change is acceptable because the Completion DISCUSSION OF CHANGES ITS 3.5.4, REFUELING WATER STORAGE TANK Sequoyah Unit 1 and Unit 2 Page 3 of 3 Time is consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. The primary function of the RWST is to provide large volumes of water to the RCS following a Loss of Coolant Accident. This large volume of water continues to be available while in this Condition. As a result, the most important safety function of the RWST can still be provided. Because of the volume of the RWST, changes to the boron concentration or temperature occur slowly, and consequently would not go far out of limit. If one of these parameters was out of limit, more than one hour would likely be required to restore the parameter. Given the remaining abilities of the RWST, requiring a plant shutdown after one hour is not warranted. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) RWST 3.5.4 Westinghouse STS 3.5.4-1 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 43.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron concentration not within limits. OR RWST borated water temperature not within limits. A.1 Restore RWST to OPERABLE status. 8 hours B. RWST inoperable for reasons other than Condition A. B.1 Restore RWST to OPERABLE status. 1 hour C. Required Action and associated Completion Time not met. C.1 Be in MODE 3.

AND C.2 Be in MODE 5. 6 hours

36 hours 3.5.5 Applicability DOC L01 ACTION ACTION RWST 3.5.4 Westinghouse STS 3.5.4-2 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 1 4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 -------------------------------NOTE------------------------------ [ Only required to be performed when ambient air temperature is < [35]°F or > [100]°F. ] --------------------------------------------------------------------- Verify RWST borated water temperature is [35]°F and [100]°F.

[ 24 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.4.2 Verify RWST borated water volume is [466,200 gallons ( )%]. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.4.3 Verify RWST boron concentration is [2000] ppm and [2200] ppm. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] 60 105 3.5.5.c 3.5.5.d 4.5.5.b 3.5.5.a 4.5.5.a.1 3.5.5.b 4.5.5.a.2 2700 2500 321333332370,000 2 RWST 3.5.4 Westinghouse STS 3.5.4-1 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 43.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron concentration not within limits. OR RWST borated water temperature not within limits. A.1 Restore RWST to OPERABLE status. 8 hours B. RWST inoperable for reasons other than Condition A. B.1 Restore RWST to OPERABLE status. 1 hour C. Required Action and associated Completion Time not met. C.1 Be in MODE 3.

AND C.2 Be in MODE 5. 6 hours

36 hours 3.5.5 Applicability DOC L01 ACTION ACTION RWST 3.5.4 Westinghouse STS 3.5.4-2 Rev. 4.0 CTS Amendment XXXSEQUOYAH UNIT 2 4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 -------------------------------NOTE------------------------------ [ Only required to be performed when ambient air temperature is < [35]°F or > [100]°F. ] --------------------------------------------------------------------- Verify RWST borated water temperature is [35]°F and [100]°F.

[ 24 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.4.2 Verify RWST borated water volume is [466,200 gallons ( )%]. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.4.3 Verify RWST boron concentration is [2000] ppm and [2200] ppm. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] 60 105 3.5.5.c 3.5.5.d 4.5.5.b 3.5.5.a 4.5.5.a.1 3.5.5.b 4.5.5.a.2 2700 2500 321333332370,000 2 JUSTIFICATION FOR DEVIATIONS ITS 3.5.4, REFUELING WATER STORAGE TANKS (RWST) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. A bracketed Note for ISTS SR 3.5.4.1 associated with the effect of ambient air temperature on RWST temperature is not adopted. SQN RWST borated water is heated and not maintained at ambient temperature, and the current temperature band is not very large.

2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. ISTS SR 3.5.4.1 (ITS SR 3.5.4.1), ISTS SR 3.5.4.2 (ITS SR 3.5.4.2), and ISTS SR 3.5.4.3 (ITS SR 3.5.4.3) provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies for these SRs under the Surveillance Frequency Control Program. 4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) RWST B 3.5.4 Westinghouse STS B 3.5.4-1 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.4 Refueling Water Storage Tank (RWST)

BASES BACKGROUND The RWST supplies borated water to the Chemical and Volume Control System (CVCS) during abnormal operating conditions, to the refueling pool during refueling, and to the ECCS and the Containment Spray System during accident conditions.

The RWST supplies both trains of the ECCS and the Containment Spray System through separate, redundant supply headers during the injection phase of a loss of coolant accident (LOCA) recovery. A motor operated isolation valve is provided in each header to isolate the RWST from the ECCS once the system has been transferred to the recirculation mode. The recirculation mode is entered when pump suction is transferred to the containment sump following receipt of the RWST - Low Low (Level 1) signal. Use of a single RWST to supply both trains of the ECCS and Containment Spray System is acceptable since the RWST is a passive component, and passive failures are not required to be assumed to occur coincidentally with Design Basis Events.

The switchover from normal operation to the injection phase of ECCS operation requires changing centrifugal charging pump suction from the CVCS volume control tank (VCT) to the RWST through the use of isolation valves. Each set of isolation valves is interlocked so that the VCT isolation valves will begin to close once the RWST isolation valves are fully open. Since the VCT is under pressure, the preferred pump suction will be from the VCT until the tank is isolated. This will result in a delay in obtaining the RWST borated water. The effects of this delay are discussed in the Applicable Safety Analyses section of these Bases. During normal operation in MODES 1, 2, and 3, the safety injection (SI) and residual heat removal (RHR) pumps are aligned to take suction from the RWST. The ECCS and Containment Spray System pumps are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at or near shutoff head conditions.

When the suction for the ECCS and Containment Spray System pumps is transferred to the containment sump, the RWST flow paths must be isolated to prevent a release of the containment sump contents to the RWST, which could result in a release of contaminants to the atmosphere and the eventual loss of suction head for the ECCS pumps. a common are s M 111cavity The areECCS INSERT 2 theINSERT 1 1 Insert Page B 3.5.4-1 INSERT 1 coincident with Containment Sump Level - High signal INSERT 2 The transfer of the containment spray pump suction to the containment sump is manually initiated upon receipt of a high level in the containment sump or the RWST Low-Low (Level) alarm. 11 RWST B 3.5.4 Westinghouse STSB 3.5.4-2Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES BACKGROUND (continued) This LCO ensures that: a.The RWST contains sufficient borated water to support the ECCSduring the injection phase, b.Sufficient water volume exists in the containment sump to supportcontinued operation of the ECCS and Containment Spray Systempumps at the time of transfer to the recirculation mode of cooling, andc.The reactor remains subcritical following a LOCA.Insufficient water in the RWST could result in insufficient cooling capacity when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment. APPLICABLE During accident conditions, the RWST provides a source of borated water SAFETY to the ECCS and Containment Spray System pumps. As such, it ANALYSES provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS - Operating," B 3.5.3, "ECCS - Shutdown," and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses. The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The importance of its value is small for units with a boron injection tank (BIT) with a high boron concentration. For units with no BIT or reduced BIT boron requirements, the minimum boron concentration limit is an important assumption in ensuring the required shutdown ;; 2213 RWST B 3.5.4 Westinghouse STSB 3.5.4-3Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES APPLICABLE SAFETY ANALYSES (continued) capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting. The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as [27] seconds, with offsite power available, or [37] seconds without offsite power. This response time includes [2] seconds for electronics delay, a [15] second stroke time for the RWST valves, and a [10] second stroke time for the VCT valves. Plants with a BIT need not be concerned with the delay since the BIT will supply highly borated water prior to RWST switchover, provided the BIT is between the pumps and the core. For a large break LOCA analysis, the minimum water volume limit of [466,200] gallons and the lower boron concentration limit of [2000] ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core. The upper limit on boron concentration of [2200] ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident. In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of [35]°F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. The upper temperature limit of [100]°F is used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower RWST temperature 2500 2700 minimize the potential for1 28 58 14441113370,000 recirculation 3RWST analysis RWST B 3.5.4 Westinghouse STS B 3.5.4-4 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

APPLICABLE SAFETY ANALYSES (continued) limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment. The RWST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode. To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs. APPLICABILITY In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A.1 With RWST boron concentration or borated water temperature not within limits, they must be returned to within limits within 8 hours. Under these conditions neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition. The 8 hour limit to restore the RWST temperature or boron concentration to within limits was developed considering the time required to change either the boron concentration or temperature and the fact that the contents of the tank are still available for injection. B.1 With the RWST inoperable for reasons other than Condition A (e.g., water volume), it must be restored to OPERABLE status within 1 hour. RWST 511 RWST B 3.5.4 Westinghouse STS B 3.5.4-5 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

ACTIONS (continued) In this Condition, neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains. C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.5.4.1 REQUIREMENTS The RWST borated water temperature should be verified to be within the limits assumed in the accident analyses band. [ The Frequency of 24 hours is sufficient to identify a temperature change that would approach either limit and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST. With ambient air temperatures within the band, the RWST temperature should not exceed the limits. RWST c 67819 RWST B 3.5.4 Westinghouse STS B 3.5.4-6 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.4.2 The RWST water volume should be verified to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. [ Since the RWST volume is normally stable and is protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.4.3 The boron concentration of the RWST should be verified to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. [ Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron concentration is appropriate and has been shown to be acceptable through operating experience.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 766 RWST B 3.5.4 Westinghouse STS B 3.5.4-7 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued)

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. FSAR, Chapter [6] and Chapter [15]. U 174 RWST B 3.5.4 Westinghouse STS B 3.5.4-1 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.4 Refueling Water Storage Tank (RWST)

BASES BACKGROUND The RWST supplies borated water to the Chemical and Volume Control System (CVCS) during abnormal operating conditions, to the refueling pool during refueling, and to the ECCS and the Containment Spray System during accident conditions.

The RWST supplies both trains of the ECCS and the Containment Spray System through separate, redundant supply headers during the injection phase of a loss of coolant accident (LOCA) recovery. A motor operated isolation valve is provided in each header to isolate the RWST from the ECCS once the system has been transferred to the recirculation mode. The recirculation mode is entered when pump suction is transferred to the containment sump following receipt of the RWST - Low Low (Level 1) signal. Use of a single RWST to supply both trains of the ECCS and Containment Spray System is acceptable since the RWST is a passive component, and passive failures are not required to be assumed to occur coincidentally with Design Basis Events.

The switchover from normal operation to the injection phase of ECCS operation requires changing centrifugal charging pump suction from the CVCS volume control tank (VCT) to the RWST through the use of isolation valves. Each set of isolation valves is interlocked so that the VCT isolation valves will begin to close once the RWST isolation valves are fully open. Since the VCT is under pressure, the preferred pump suction will be from the VCT until the tank is isolated. This will result in a delay in obtaining the RWST borated water. The effects of this delay are discussed in the Applicable Safety Analyses section of these Bases. During normal operation in MODES 1, 2, and 3, the safety injection (SI) and residual heat removal (RHR) pumps are aligned to take suction from the RWST. The ECCS and Containment Spray System pumps are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at or near shutoff head conditions.

When the suction for the ECCS and Containment Spray System pumps is transferred to the containment sump, the RWST flow paths must be isolated to prevent a release of the containment sump contents to the RWST, which could result in a release of contaminants to the atmosphere and the eventual loss of suction head for the ECCS pumps. a common are s M 111cavity The areECCS INSERT 2 theINSERT 1 1 Insert Page B 3.5.4-1 INSERT 1 coincident with Containment Sump Level - High signal

INSERT 2 The transfer of the containment spray pump suction to the containment sump is manually initiated upon receipt of a high level in the containment sump or the RWST Low-Low (Level) alarm. 11 RWST B 3.5.4 Westinghouse STSB 3.5.4-2Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES BACKGROUND (continued) This LCO ensures that: a.The RWST contains sufficient borated water to support the ECCSduring the injection phase, b.Sufficient water volume exists in the containment sump to supportcontinued operation of the ECCS and Containment Spray Systempumps at the time of transfer to the recirculation mode of cooling, andc.The reactor remains subcritical following a LOCA.Insufficient water in the RWST could result in insufficient cooling capacity when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment. APPLICABLE During accident conditions, the RWST provides a source of borated water SAFETY to the ECCS and Containment Spray System pumps. As such, it ANALYSES provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS - Operating," B 3.5.3, "ECCS - Shutdown," and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses. The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The importance of its value is small for units with a boron injection tank (BIT) with a high boron concentration. For units with no BIT or reduced BIT boron requirements, the minimum boron concentration limit is an important assumption in ensuring the required shutdown ;; 2213 RWST B 3.5.4 Westinghouse STSB 3.5.4-3Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES APPLICABLE SAFETY ANALYSES (continued) capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting. The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as [27] seconds, with offsite power available, or [37] seconds without offsite power. This response time includes [2] seconds for electronics delay, a [15] second stroke time for the RWST valves, and a [10] second stroke time for the VCT valves. Plants with a BIT need not be concerned with the delay since the BIT will supply highly borated water prior to RWST switchover, provided the BIT is between the pumps and the core. For a large break LOCA analysis, the minimum water volume limit of [466,200] gallons and the lower boron concentration limit of [2000] ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core. The upper limit on boron concentration of [2200] ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident. In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of [35]°F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. The upper temperature limit of [100]°F is used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower RWST temperature 2500 2700 minimize the potential for1 28 58 14441113370,000 recirculation 3RWST analysis RWST B 3.5.4 Westinghouse STS B 3.5.4-4 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

APPLICABLE SAFETY ANALYSES (continued) limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment. The RWST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode. To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs. APPLICABILITY In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." ACTIONS A.1 With RWST boron concentration or borated water temperature not within limits, they must be returned to within limits within 8 hours. Under these conditions neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition. The 8 hour limit to restore the RWST temperature or boron concentration to within limits was developed considering the time required to change either the boron concentration or temperature and the fact that the contents of the tank are still available for injection. B.1 With the RWST inoperable for reasons other than Condition A (e.g., water volume), it must be restored to OPERABLE status within 1 hour. RWST 511 RWST B 3.5.4 Westinghouse STS B 3.5.4-5 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

ACTIONS (continued) In this Condition, neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains. C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.5.4.1 REQUIREMENTS The RWST borated water temperature should be verified to be within the limits assumed in the accident analyses band. [ The Frequency of 24 hours is sufficient to identify a temperature change that would approach either limit and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST. With ambient air temperatures within the band, the RWST temperature should not exceed the limits. RWST c 67819 RWST B 3.5.4 Westinghouse STS B 3.5.4-6 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued) SR 3.5.4.2 The RWST water volume should be verified to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. [ Since the RWST volume is normally stable and is protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.4.3 The boron concentration of the RWST should be verified to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. [ Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron concentration is appropriate and has been shown to be acceptable through operating experience.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 766 RWST B 3.5.4 Westinghouse STS B 3.5.4-7 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX1BASES

SURVEILLANCE REQUIREMENTS (continued)

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. FSAR, Chapter [6] and Chapter [15]. U 174 JUSTIFICATION FOR DEVIATIONS ITS 3.5.4 BASES, REFUELING WATER STORAGE TANK (RWST) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. 2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3. 3. ISTS Bases 3.5.4 Applicable Safety Analyses section, second paragraph, includes the phrase "The importance of its value is small for units with a boron injection tank (BIT) with a high boron concentration. For units with no BIT or reduced BIT boron requirements..." ISTS Bases 3.5.4 Applicable Safety Analyses section, third paragraph, includes the sentence "Plants with a BIT need not be concerned with the delay since the BIT will supply highly borated water prior to RWST switchover, provided the BIT is between the pumps and the core." ITS Bases 3.5.4 does not include this phrase and sentence. This is acceptable, because the plant specific design includes a BIT, but there are no minimum boron concentration design and licensing basis requirements for the BIT. Therefore, deletion of this information is consistent with the current plant design and licensing basis. 4. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 5. The MODE 5 and MODE 6 core cooling LCOs listed are relative to the shutdown cooling function of the RHR System and not the ECCS function. The Applicability Section has adequately described why ECCS is not needed in MODES 5 and 6, and it is not necessary to address the requirements of the shutdown cooling function. Therefore, this inappropriate information has been deleted. 6. ISTS SR 3.5.4.1, ISTS SR 3.5.4.2, and ITS SR 3.5.4.3 provide two options for controlling the Frequencies of Surveillance Requirements. SQN is proposing to control the Surveillance Frequencies under the Surveillance Frequency Control Program. Therefore, the Frequency for ITS SR 3.5.4.1, SR 3.5.4.2, and SR 3.5.4.3 is "In accordance with the Surveillance Frequency Control Program." 7. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 8. Changes are made to reflect those changes made to the Specification. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.

9. Typographical/grammatical error corrected.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.5.4, REFUELING WATER STORAGE TANK Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 5 ITS 3.5.5, SEAL INJECTION FLOW Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 3.5.5COOLING SYSTEMS (ECCS) 3/4.5.6 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.6 Reactor coolant pump seal injection flow shall be within the limits of Figure 3.5.6-1.

APPLICABILITY: MODES 1, 2, and 3. ACTION: With reactor coolant pump seal injection flow not within limits, adjust manual seal injection throttle valves to give a flow within limit in accordance with Surveillance Requirement 4.5.6 within 4 hours. Otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.5.6 At least once per 31 days* verify manual seal injection throttle valves are adjusted to give a flow within the emergency core cooling system safety analysis limits in Figure 3.5.6-1.

    *This surveillance is not required to be performed until 4 hours after the reactor coolant system pressure stabilizes at  2215 psig and  2255 psig.

January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-8 Amendment No. 259, 326 LCO 3.5.5 Applicability ACTION B ACTION A SR 3.5.5.1 SR 3.5.5.1 Note LA01In accordance with the Surveillance Frequency Control Program Page 1 of 4 A01ITS ITS 3.5.5 FIGURE 3.5.6-1

January 28, 2010 Sequoyah - Unit 1 3/4 5-9 Amendment No. 259, 326 Page 2 of 4 Figure 3.5.5-1 A01ITS ITS 3.5.5EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.6 SEAL INJECTION FLOW

LIMITING CONDITION FOR OPERATION

3.5.6 Reactor coolant pump seal injection flow shall be within the limits of Figure 3.5.6-1.

APPLICABILITY: MODES 1, 2, and 3.

ACTION: With reactor coolant pump seal injection flow not within limits, adjust manual seal injection throttle valves to give a flow within limit in accordance with Surveillance Requirement 4.5.6 within 4 hours. Otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

SURVEILLANCE REQUIREMENTS

4.5.6 At least once per 31 days* verify manual seal injection throttle valves are adjusted to give a flow within the emergency core cooling system safety analysis limits in Figure 3.5.6-1.

This surveillance is not required to be performed until 4 hours after the reactor coolant system pressure stabilizes at 2215 psig and 2255 psig.

January 28, 2010 SEQUOYAH - UNIT 2 3/4 5-8 Amendment No. 250, 319 LCO 3.5.5 Applicability ACTION A ACTION B SR 3.5.5.1 LA01In accordance with the Surveillance Frequency Control Program SR 3.5.5.1 Note Page 3 of 4 A01ITS ITS 3.5.5 FIGURE 3.5.6-1

January 28, 2010 SEQUOYAH UNIT 2 3/4 5-9 Amendment No. 250, 319 Page 4 of 4 Figure 3.5.5-1 DISCUSSION OF CHANGES ITS 3.5.5, SEAL INJECTION FLOW Sequoyah Unit 1 and Unit 2 Page 1 of 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.5.6 requires, in part, verifying manual seal injection throttle valve are adjusted to give a flow within the limits of Figure 3.5.6-1 every 31 days. ITS SR 3.5.5.1 requires a similar Surveillance and specifies a periodic Frequency of, "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequency for this SR and associated Bases to the Surveillance Frequency Control Program. The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable, because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Surveillance Frequency is removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 3.5.5, SEAL INJECTION FLOW Sequoyah Unit 1 and Unit 2 Page 2 of 2 LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-1 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 1 CTS 33.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow

LCO 3.5.5 Reactor coolant pump seal injection flow [resistance] shall be [ [40] gpm with [centrifugal charging pump discharge header] pressure [2480] psig and the [charging flow] control valve full open or [0.2117] ft/gpm2 or within the limits of Figure 3.5.5-1].

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow [resistance] not within limit. A.1 Adjust manual seal injection throttle valves to give a flow [resistance] within limit. 4 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 4. 6 hours

12 hours

3.5.6 Applicability ACTION ACTION 11 Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-2 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 1 CTS 3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 -------------------------------NOTE------------------------------ Not required to be performed until 4 hours after the Reactor Coolant System pressure stabilizes at [2215 psig and 2255 psig]. ---------------------------------------------------------------------

Verify manual seal injection throttle valves are adjusted to give a flow [resistance] [of [40 gpm] with [centrifugal charging pump discharge header] pressure [2480] psig and the [charging flow] control valve full open or [0.2117] ft/gpm2 or within the limit of Figure 3.5.5-1.]

[ 31 days OR In accordance with the Surveillance Frequency Control Program ] 4.5.6 4.5.6

1221 Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-3 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 1 CTS 3

Figure 3.5.5-1 (page 1 of 1) Seal Injection Flow Limits For illustration only. Do not use for operation. INSERT 1 3 3.5.5 Insert Page 3.5.5-3 CTS INSERT 1

3Figure 3.5.6-1 Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-1 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 2 CTS 33.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow

LCO 3.5.5 Reactor coolant pump seal injection flow [resistance] shall be [ [40] gpm with [centrifugal charging pump discharge header] pressure [2480] psig and the [charging flow] control valve full open or [0.2117] ft/gpm2 or within the limits of Figure 3.5.5-1].

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow [resistance] not within limit. A.1 Adjust manual seal injection throttle valves to give a flow [resistance] within limit. 4 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE 3.

AND B.2 Be in MODE 4. 6 hours

12 hours

3.5.6 Applicability ACTION ACTION 11 Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-2 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 2 CTS 3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 -------------------------------NOTE------------------------------ Not required to be performed until 4 hours after the Reactor Coolant System pressure stabilizes at [2215 psig and 2255 psig]. ---------------------------------------------------------------------

Verify manual seal injection throttle valves are adjusted to give a flow [resistance] [of [40 gpm] with [centrifugal charging pump discharge header] pressure [2480] psig and the [charging flow] control valve full open or [0.2117] ft/gpm2 or within the limit of Figure 3.5.5-1.]

[ 31 days OR In accordance with the Surveillance Frequency Control Program ] 4.5.6 4.5.6

1221 Seal Injection Flow 3.5.5 Westinghouse STS 3.5.5-3 Rev. 4.0 Amendment XXXSEQUOYAH UNIT 2 CTS 3

Figure 3.5.5-1 (page 1 of 1) Seal Injection Flow Limits For illustration only. Do not use for operation. INSERT 1 3 3.5.5 Insert Page 3.5.5-3 CTS INSERT 1

3Figure 3.5.6-1 JUSTIFICATION FOR DEVIATIONS ITS 3.5.5, SEAL INJECTION FLOW Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 2. ISTS SR 3.5.5.1 (ITS SR 3.5.5.1) provide two options for controlling the Frequency of the Surveillance Requirement. SQN is proposing to control the Surveillance Frequency for this SR under the Surveillance Frequency Control Program. 3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-1 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX2B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.5 Seal Injection Flow

BASES BACKGROUND This LCO is applicable only to those units that utilize the centrifugal charging pumps for safety injection (SI). The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS). The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident. This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during SI. The RCP seal injection flow is restricted by the seal injection line flow [resistance] which is adjusted through positioning of the manual RCP seal injection throttle valves. The RCP seal injection flow [resistance] is determined by measuring the pressurizer pressure, the centrifugal charging pump discharge header pressure, and the RCP seal injection flow rate. The charging flow control valve throttles the centrifugal charging pump discharge header flow as necessary to maintain the programmed level in the pressurizer. The charging flow control valve fails open to ensure that, in the event of either loss of air or loss of control signal to the valve, when the centrifugal charging pumps are supplying charging flow, seal injection flow to the RCP seals is maintained. Positioning of the charging flow control valve may vary during normal plant operating conditions, resulting in a proportional change to RCP seal injection flow. The flow [resistance] provided by RCP seal injection throttle valves will remain fixed when the charging flow control valve is repositioned provided the throttle valve(s) position are not adjusted. APPLICABLE All ECCS subsystems are taken credit for in the large break loss of SAFETY coolant accident (LOCA) at full power (Ref. 1). The LOCA analysis ANALYSES establishes the minimum flow for the ECCS pumps. The centrifugal charging pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps. The steam generator tube rupture and main steam line break event analyses also credit the centrifugal charging pumps, but are not limiting in their design. Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses. 3312safety injection (SI) Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-2 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX2BASES

APPLICABLE SAFETY ANALYSES (continued) This LCO ensures that seal injection flow [resistance] will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA. It also ensures that the centrifugal charging pumps will deliver sufficient water for a small LOCA and sufficient boron to maintain the core subcritical. For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory. Seal injection flow satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2). [ The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance. In order to establish the proper flow line resistance, a pressure and flow must be known. The flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure and that the centrifugal charging pump discharge pressure is greater than or equal to the value specified in this LCO. The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO. A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure. The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The additional modifier of this LCO, the control valve (charging flow for four loop units and air operated seal injection for three loop units) being full open, is required since the valve is designed to fail open for the accident condition. With the discharge pressure and control valve position as specified by the LCO, a flow limit is established. It is this flow limit that is used in the accident analyses. OR This is accomplished by limiting the seal injection line resistance to a value consistent with the assumptions in the accident analysis. The limit on RCP seal injection flow resistance must be met to assure that the ECCS is OPERABLE. If this limit is not met, the ECCS flow may not be as assumed in the accident analysis. The restriction on seal injection flow is accomplished by maintaining the seal water injection flow resistance [0.2117] ft/gpm2. With the seal injection flow resistances within limit, the resulting total seal injection flow will be within the assumptions made for seal flow during accident conditions. 33 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-3 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX2BASES

LCO (continued) In order to establish the proper flow line resistance, the centrifugal charging pump discharge header pressure, the RCP seal injection flow rate, and the pressurizer pressure are measured. The line resistance is then determined from those inputs. A reduction in RCP pressure with no concurrent decrease in centrifugal charging pump discharge header pressure would increase the differential pressure across the manual throttle valves, and result in more flow being discharged through the RCP seal injection line. The flow resistance limit assures that when RCS pressure drops during a LOCA and seal injection flow increases in response to the higher differential pressure, the resulting flow will be consistent with the accident analysis. OR The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance. In order to establish the proper flow line resistance, a pressure and flow must be known. The flow line resistance is established by adjusting the RCP seal injection flow in the acceptable region of Figure 3.5.5-1 at a given pressure differential between the charging header and the RCS. The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO. A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure. The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The flow limits established by Figure 3.5.5-1 ensures that the minimum ECCS flow assumed in the safety analyses is maintained.] The limit on seal injection flow [resistance] must be met to render the ECCS OPERABLE. If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses. APPLICABILITY In MODES 1, 2, and 3, the seal injection flow [resistance] limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4. The seal injection flow [resistance] limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow [resistance] must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance. 333333seal injection throttle valves (needle valves) to provide a total 2 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-4 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX2BASES

ACTIONS A.1 With the seal injection flow [resistance] not within its limit, the amount of charging flow available to the RCS may be reduced. Under this Condition, action must be taken to restore the flow [resistance] to within its limit. The operator has 4 hours from the time the flow [resistance] is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis. The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow [resistance] within limits. This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable. SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification that the manual seal injection throttle valves are adjusted to give a flow [resistance] within the limit ensures that the ECCS injection flows stay within the safety analysis. A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to within the limit determined in accordance with the ECCS safety analysis. [The flow [resistance] shall be verified by confirming seal injection flow [40] gpm with the RCS at normal operating pressure, the charging flow control valve full open, and the charging header pressure [2480]. OR The flow [resistance] shall be verified by confirming seal injection flow and differential pressure within the acceptable region of Figure 3.5.5-1. 33333 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-5 Rev. 4.0 SEQUOYAH UNIT 1 Revision XXX2BASES

SURVEILLANCE REQUIREMENTS (continued) OR The flow resistance shall be [0.2117] ft/gpm2.] Control valves in the flow path between the charging header and the RCS pressure sensing points must be in their post accident position (e.g., charging flow control valve open) during this Surveillance to correlate with the acceptance criteria. [ The Frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve Surveillance Frequencies. The Frequency has proven to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] As noted, the Surveillance is not required to be performed until 4 hours after the RCS pressure has stabilized within a +/- 20 psig range of normal operating pressure. The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly. The exception is limited to 4 hours to ensure that the Surveillance is timely. REFERENCES 1. FSAR, Chapter [6] and Chapter [15].

2. 10 CFR 50.46. at 2215 psig and 2255 psig U 5422343 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-1 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX2B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.5 Seal Injection Flow

BASES BACKGROUND This LCO is applicable only to those units that utilize the centrifugal charging pumps for safety injection (SI). The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS). The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident. This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during SI. The RCP seal injection flow is restricted by the seal injection line flow [resistance] which is adjusted through positioning of the manual RCP seal injection throttle valves. The RCP seal injection flow [resistance] is determined by measuring the pressurizer pressure, the centrifugal charging pump discharge header pressure, and the RCP seal injection flow rate. The charging flow control valve throttles the centrifugal charging pump discharge header flow as necessary to maintain the programmed level in the pressurizer. The charging flow control valve fails open to ensure that, in the event of either loss of air or loss of control signal to the valve, when the centrifugal charging pumps are supplying charging flow, seal injection flow to the RCP seals is maintained. Positioning of the charging flow control valve may vary during normal plant operating conditions, resulting in a proportional change to RCP seal injection flow. The flow [resistance] provided by RCP seal injection throttle valves will remain fixed when the charging flow control valve is repositioned provided the throttle valve(s) position are not adjusted. APPLICABLE All ECCS subsystems are taken credit for in the large break loss of SAFETY coolant accident (LOCA) at full power (Ref. 1). The LOCA analysis ANALYSES establishes the minimum flow for the ECCS pumps. The centrifugal charging pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps. The steam generator tube rupture and main steam line break event analyses also credit the centrifugal charging pumps, but are not limiting in their design. Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses. 3312safety injection (SI) Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-2 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX2BASES

APPLICABLE SAFETY ANALYSES (continued) This LCO ensures that seal injection flow [resistance] will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA. It also ensures that the centrifugal charging pumps will deliver sufficient water for a small LOCA and sufficient boron to maintain the core subcritical. For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory. Seal injection flow satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2). [ The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance. In order to establish the proper flow line resistance, a pressure and flow must be known. The flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure and that the centrifugal charging pump discharge pressure is greater than or equal to the value specified in this LCO. The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO. A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure. The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The additional modifier of this LCO, the control valve (charging flow for four loop units and air operated seal injection for three loop units) being full open, is required since the valve is designed to fail open for the accident condition. With the discharge pressure and control valve position as specified by the LCO, a flow limit is established. It is this flow limit that is used in the accident analyses. OR This is accomplished by limiting the seal injection line resistance to a value consistent with the assumptions in the accident analysis. The limit on RCP seal injection flow resistance must be met to assure that the ECCS is OPERABLE. If this limit is not met, the ECCS flow may not be as assumed in the accident analysis. The restriction on seal injection flow is accomplished by maintaining the seal water injection flow resistance [0.2117] ft/gpm2. With the seal injection flow resistances within limit, the resulting total seal injection flow will be within the assumptions made for seal flow during accident conditions. 33 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-3 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX2BASES

LCO (continued) In order to establish the proper flow line resistance, the centrifugal charging pump discharge header pressure, the RCP seal injection flow rate, and the pressurizer pressure are measured. The line resistance is then determined from those inputs. A reduction in RCP pressure with no concurrent decrease in centrifugal charging pump discharge header pressure would increase the differential pressure across the manual throttle valves, and result in more flow being discharged through the RCP seal injection line. The flow resistance limit assures that when RCS pressure drops during a LOCA and seal injection flow increases in response to the higher differential pressure, the resulting flow will be consistent with the accident analysis. OR The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance. In order to establish the proper flow line resistance, a pressure and flow must be known. The flow line resistance is established by adjusting the RCP seal injection flow in the acceptable region of Figure 3.5.5-1 at a given pressure differential between the charging header and the RCS. The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO. A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure. The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The flow limits established by Figure 3.5.5-1 ensures that the minimum ECCS flow assumed in the safety analyses is maintained.] The limit on seal injection flow [resistance] must be met to render the ECCS OPERABLE. If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses. APPLICABILITY In MODES 1, 2, and 3, the seal injection flow [resistance] limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4. The seal injection flow [resistance] limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow [resistance] must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance. 333333seal injection throttle valves (needle valves) to provide a total 2 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-4 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX2BASES

ACTIONS A.1 With the seal injection flow [resistance] not within its limit, the amount of charging flow available to the RCS may be reduced. Under this Condition, action must be taken to restore the flow [resistance] to within its limit. The operator has 4 hours from the time the flow [resistance] is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis. The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow [resistance] within limits. This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable. SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification that the manual seal injection throttle valves are adjusted to give a flow [resistance] within the limit ensures that the ECCS injection flows stay within the safety analysis. A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to within the limit determined in accordance with the ECCS safety analysis. [The flow [resistance] shall be verified by confirming seal injection flow [40] gpm with the RCS at normal operating pressure, the charging flow control valve full open, and the charging header pressure [2480]. OR The flow [resistance] shall be verified by confirming seal injection flow and differential pressure within the acceptable region of Figure 3.5.5-1. 33333 Seal Injection Flow B 3.5.5 Westinghouse STS B 3.5.5-5 Rev. 4.0 SEQUOYAH UNIT 2 Revision XXX2BASES

SURVEILLANCE REQUIREMENTS (continued) OR The flow resistance shall be [0.2117] ft/gpm2.] Control valves in the flow path between the charging header and the RCS pressure sensing points must be in their post accident position (e.g., charging flow control valve open) during this Surveillance to correlate with the acceptance criteria. [ The Frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve Surveillance Frequencies. The Frequency has proven to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] As noted, the Surveillance is not required to be performed until 4 hours after the RCS pressure has stabilized within a +/- 20 psig range of normal operating pressure. The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly. The exception is limited to 4 hours to ensure that the Surveillance is timely. REFERENCES 1. FSAR, Chapter [6] and Chapter [15].

2. 10 CFR 50.46. at 2215 psig and 2255 psig U 5422343 JUSTIFICATION FOR DEVIATIONS ITS 3.5.5 BASES, SEAL WATER INJECTION Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The deleted sentence is not required based on SQN utilizes centrifugal charging pumps for safety injection. 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. 3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 4. ISTS SR 3.5.1.1 Bases provide two options for controlling the Frequency of the Surveillance Requirement. SQN is proposing to control the Surveillance Frequency for ITS SR 3.5.1.1 under the Surveillance Frequency Control Program. 5. The Reviewer's Note has been deleted, because it is not meant to be retained in the plant specific ITS submittal.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.5.5, SEAL INJECTION FLOW Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 6 Improved Standard Technical Specifications (ISTS) Not Adopted in the Sequoyah ITS ISTS 3.5.6, BORON INJECTION TANK (BIT) BIT 3.5.6 Westinghouse STS 3.5.6-1 Rev. 4.0 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Boron Injection Tank (BIT) LCO 3.5.6 The BIT shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. BIT inoperable. A.1 Restore BIT to OPERABLE status. 1 hour B. Required Action and associated Completion Time of Condition A not met. B.1 Be in MODE 3. AND B.2 Borate to SDM specified in COLR. AND B.3 Restore BIT to OPERABLE status. 6 hours 6 hours

7 days C. Required Action and associated Completion Time of Condition B not met. C.1 Be in MODE 4. 12 hours

1 BIT 3.5.6 Westinghouse STS 3.5.6-2 Rev. 4.0 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.6.1 Verify BIT borated water temperature is [145]°F. [ 24 hours OR In accordance with the Surveillance Frequency Control Program ] SR 3.5.6.2 [ Verify BIT borated water volume is [1100] gallons. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] ] SR 3.5.6.3 Verify BIT boron concentration is [20,000] ppm and [22,500] ppm. [ 7 days OR In accordance with the Surveillance Frequency Control Program ] 1 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) JUSTIFICATION FOR DEVIATIONS ISTS 3.5.6, BORON INJECTION TANK (BIT) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. ISTS 3.5.6, Boron Injection Tank (BIT) is not being adopted because Sequoyah Nuclear Plant (SQN) design does not include the BIT. Therefore, ISTS 3.5.6 is not included in the ITS. . Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs) BIT B 3.5.6 Westinghouse STS B 3.5.6-1 Rev. 4.0 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.6 Boron Injection Tank (BIT) BASES BACKGROUND The BIT is part of the Boron Injection System, which is the primary means of quickly introducing negative reactivity into the Reactor Coolant System (RCS) on a safety injection (SI) signal. The main flow path through the Boron Injection System is from the discharge of the centrifugal charging pumps through lines equipped with a flow element and two valves in parallel that open on an SI signal. The valves can be operated from the main control board. The valves and flow elements have main control board indications. Downstream of these valves, the flow enters the BIT (Ref. 1). The BIT is a stainless steel tank containing concentrated boric acid. Two trains of strip heaters are mounted on the tank to keep the temperature of the boric acid solution above the precipitation point. The strip heaters are controlled by temperature elements located near the bottom of the BIT. The temperature elements also activate High and Low alarms on the main control board. In addition to the strip heaters on the BIT, there is a recirculation system with a heat tracing system, including the piping section between the motor operated isolation valves, which further ensures that the boric acid stays in solution. The BIT is also equipped with a High Pressure alarm on the main control board. The entire contents of the BIT are injected when required; thus, the contained and deliverable volumes are the same. During normal operation, one of the two BIT recirculation pumps takes suction from the boron injection surge tank (BIST) and discharges to the BIT. The solution then returns to the BIST. Normally, one pump is running and one is shut off. On receipt of an SI signal, the running pump shuts off and the air operated valves close. Flow to the BIT is then supplied from the centrifugal charging pumps. The solution of the BIT is injected into the RCS through the RCS cold legs. APPLICABLE During a main steam line break (MSLB) or loss of coolant accident SAFETY (LOCA), the BIT provides an immediate source of concentrated boric ANALYSES acid that quickly introduces negative reactivity into the RCS. The contents of the BIT are not credited for core cooling or immediate boration in the LOCA analysis, but for post LOCA recovery. The BIT maximum boron concentration of [22,500] ppm is used to determine the minimum time for hot leg recirculation switchover. The minimum boron 1 BIT B 3.5.6 Westinghouse STS B 3.5.6-2 Rev. 4.0 BASES APPLICABLE SAFETY ANALYSES (continued) concentration of [20,000] ppm is used to determine the minimum mixed mean sump boron concentration for post LOCA shutdown requirements. For the MSLB analysis, the BIT is the primary mechanism for injecting boron into the core to counteract any positive increases in reactivity caused by an RCS cooldown. The analysis uses the minimum boron concentration of the BIT, which also affects both the departure from nucleate boiling and containment design analyses. Reference to the LOCA and MSLB analyses is used to assess changes to the BIT to evaluate their effect on the acceptance limits contained in these analyses. The minimum temperature limit of [145]°F for the BIT ensures that the solution does not reach the boric acid precipitation point. The temperature of the solution is monitored and alarmed on the main control board. The BIT boron concentration limits are established to ensure that the core remains subcritical during post LOCA recovery. The BIT will counteract any positive increases in reactivity caused by an RCS cooldown. The BIT minimum water volume limit of [1100] gallons is used to ensure that the appropriate quantity of highly borated water with sufficient negative reactivity is injected into the RCS to shut down the core following an MSLB, to determine the hot leg recirculation switchover time, and to safeguard against boron precipitation. The BIT satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). LCO This LCO establishes the minimum requirements for contained volume, boron concentration, and temperature of the BIT inventory (Ref. 2). This ensures that an adequate supply of borated water is available in the event of a LOCA or MSLB to maintain the reactor subcritical following these accidents. To be considered OPERABLE, the limits established in the SR for water volume, boron concentration, and temperature must be met. If the equipment used to verify BIT parameters (temperature, volume, and boron concentration) is determined to be inoperable, then the BIT is also inoperable. 1 BIT B 3.5.6 Westinghouse STS B 3.5.6-3 Rev. 4.0 BASES APPLICABILITY In MODES 1, 2, and 3, the BIT OPERABILITY requirements are consistent with those of LCO 3.5.2, "ECCS - Operating." In MODES 4, 5, and 6, the respective accidents are less severe, so the BIT is not required in these lower MODES. ACTIONS A.1 If the required volume is not present in the BIT, both the hot leg recirculation switchover time analysis and the boron precipitation analysis would not be met. Under these conditions, prompt action must be taken to restore the volume to above its required limit to declare the tank OPERABLE, or the plant must be placed in a MODE in which the BIT is not required. The BIT boron concentration is considered in the hot leg recirculation switchover time analysis, the boron precipitation analysis, and the reactivity analysis for an MSLB. If the concentration were not within the required limits, these analyses could not be relied on. Under these conditions, prompt action must be taken to restore the concentration to within its required limits, or the plant must be placed in a MODE in which the BIT is not required. The BIT temperature limit is established to ensure that the solution does not reach the boric acid crystallization point. If the temperature of the solution drops below the minimum, prompt action must be taken to raise the temperature and declare the tank OPERABLE, or the plant must be placed in a MODE in which the BIT is not required. The 1 hour Completion Time to restore the BIT to OPERABLE status is consistent with other Completion Times established for loss of a safety function and ensures that the plant will not operate for long periods outside of the safety analyses. B.1, B.2, and B.3 When Required Action A.1 cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Six hours is a reasonable time, based on operating experience, to reach MODE 3 from full power conditions and to be borated to the required SDM without challenging plant systems or operators. Borating to the required SDM assures that the plant is in a safe condition, without need for any additional boration. 1 BIT B 3.5.6 Westinghouse STS B 3.5.6-4 Rev. 4.0 BASES ACTIONS (continued) After determining that the BIT is inoperable and the Required Actions of B.1 and B.2 have been completed, the tank must be returned to OPERABLE status within 7 days. These actions ensure that the plant will not be operated with an inoperable BIT for a lengthy period of time. It should be noted, however, that changes to applicable MODES cannot be made until the BIT is restored to OPERABLE status pursuant to the provisions of LCO 3.0.4.

C.1 Even though the RCS has been borated to a safe and stable condition as a result of Required Action B.2, either the BIT must be restored to OPERABLE status (Required Action C.1) or the plant must be placed in a condition in which the BIT is not required (MODE 4). The 12 hour Completion Time to reach MODE 4 is reasonable, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. SURVEILLANCE SR 3.5.6.1 REQUIREMENTS Verification that the BIT water temperature is at or above the specified minimum temperature will identify a temperature change that would approach the acceptable limit. The solution temperature is also monitored by an alarm that provides further assurance of protection against low temperature. [ The Frequency of 24 hours has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. -----------------------------------REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] 1 BIT B 3.5.6 Westinghouse STS B 3.5.6-5 Rev. 4.0 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.5.6.2 Verification that the BIT contained volume is above the required limit assures that this volume will be available for quick injection into the RCS. If the volume is too low, the BIT would not provide enough borated water to ensure subcriticality during recirculation or to shut down the core following an MSLB. [ Since the BIT volume is normally stable, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ] SR 3.5.6.3 Verification that the boron concentration of the BIT is within the required band ensures that the reactor remains subcritical following a LOCA; it limits return to power following an MSLB, and maintains the resulting sump pH in an acceptable range so that boron precipitation will not occur in the core. In addition, the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.

The BIT is in a recirculation loop that provides continuous circulation of the boric acid solution through the BIT and the boric acid tank (BAT). There are a number of points along the recirculation loop where local samples can be taken. The actual location used to take a sample of the solution is specified in the plant Surveillance procedures. Sampling from the BAT to verify the concentration of the BIT is not recommended, since this sample may not be homogenous and the boron concentration of the two tanks may differ. The sample should be taken from the BIT or from a point in the flow path of the BIT recirculation loop. 1 BIT B 3.5.6 Westinghouse STS B 3.5.6-6 Rev. 4.0 BASES SURVEILLANCE REQUIREMENTS (continued) [ The Frequency of 7 days is appropriate and has been shown to be acceptable through operating experience. OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER'S NOTE----------------------------------- Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. ------------------------------------------------------------------------------------------------ ]

REFERENCES 1. FSAR, Chapter [6] and Chapter [15]. 2. 10 CFR 50.46. 1 JUSTIFICATION FOR DEVIATIONS ITS 3.5.6 BASES, BORON INJECTION TANK (BIT) Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. ISTS 3.5.6 Bases, "Boron Injection Tank (BIT)" is not included in the Sequoyah Nuclear Plant (SQN) ITS since the Specification, ISTS 3.5.6, has not been included in the SQN ITS.

ENCLOSURE 2 VOLUME 15 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 4.0 DESIGN FEATURES Revision 0 LIST OF ATTACHMENTS 1. ITS Chapter 4.0 - DESIGN FEATURES ATTACHMENT 1 ITS 4.0, DESIGN FEATURES Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS Chapter 4.0 5.0 DESIGN FEATURES 5.1 SITE LOCATION The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant. EXCLUSION AREA 5.2.1 DELETED LOW POPULATION ZONE 5.1.2 DELETED SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 DELETED SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 DELETED 5.2 CONTAINMENT CONFIGURATION 5.2.1 DELETED DESIGN PRESSURE AND TEMPERATURE 5.2.2 DELETED

August 2, 2006 SEQUOYAH - UNIT 1 5-1 Amendment No. 309 Page 1 of 16 4.1 A01ITS ITS Chapter 4.0

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August 2, 2006 SEQUOYAH - UNIT 1 5-2 Amendment No. 42, 114, 309 Page 2 of 16 A01ITS ITS Chapter 4.0

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August 2, 2006 SEQUOYAH - UNIT 1 5-3 Amendment No. 309 Page 3 of 16 A01ITS ITS Chapter 4.0 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 DELETED VOLUME 5.4.2 DELETED

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 DELETED

August 2, 2006 SEQUOYAH - UNIT 1 5-4 Amendment No. 45, 144, 180, 258, 268, 309 4.2 4.2.1 4.2.2 Page 4 of 16 A01ITS ITS Chapter 4.0 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed for fuel enriched to 5 weight percent U-235 and shall be maintained with: a. A keff less than critical when flooded with unborated water and a keff less than or equal to 0.95 when flooded with water containing 300 ppm soluble boron.* b. A nominal 8.972 inch center-to-center distance between fuel assemblies placed in the storage racks.

c. Arrangements of one or more of three different arrays (Regions) or sub-arrays as illustrated in Figures 5.6-1 and 5.6-1a. These arrangements in the spent fuel storage pool have the following definitions:

1. Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.95 +/- 0.05 wt% U-235, (or spent fuel regardless of the fuel burnup), in a 1-in-4 checkerboard arrangement of 1 fresh assembly with 3 spent fuel assemblies with enrichment-burnup and cooling times illustrated in Figure 5.6-2 and defined by the equations in Table 5.6-1. Cooling time is defined as the period since reactor shutdown at the end of the last operating cycle for the discharged spent fuel assembly. The presence of a removable, non-fissile insert such as a burnable poison rod assembly (BPRA) or either gadolinia or integral fuel burnable absorber (IFBA) in a fresh fuel assembly does not affect the applicability of Figure 5.6-2 or Table 5.6-1. Two alternative storage arrays (or sub-arrays) are acceptable in Region 1 if the fresh fuel assemblies contain rods with either gadolinia or integral fuel burnable absorber (IFBA). For these types of assemblies, the minimum burnup of the spent fuel in the 1-of-4 sub-array are defined by the equations in Table 5.6-2. Restrictions in Region 1 Any of the three sub-arrays illustrated in Figure 5.6-1a may be used in any combination provided that: 1) Each sub-array of 4 fuel assemblies includes, in addition to the fresh fuel assembly, 3 assemblies with enrichment and minimum burnup requirements defined by the equations in Tables 5.6-1 and 5.6-2, as appropriate. 2) The arrangement of Region 1 sub-arrays must not allow a configuration with fresh assemblies adjacent to each other. 3) If Region 1 arrays are used in conjunction with Region 2 or Region 3 arrangements (see below), the arrangements shall not allow fresh fuel assemblies to be adjacent to each other (see also Figure 5.6-1). *For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident. December 19, 2000 SEQUOYAH - UNIT 1 5-5 Amendment No. 13, 60, 114, 144, 167, 2654.3 4.3.1 4.3.1.1 4.3.1.1.a 4.3.1.1.b 4.3.1.1.c 4.3.1.1.b See ITS 3.7.15 Page 5 of 16 A01ITS ITS Chapter 4.0 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed for fuel enriched to 5.0 weight percent U-235 and shall be maintained with the arrangement of 146 storage locations shown in Figure 5.6-4. The cells shown as empty cells in Figure 5.6-4 shall have physical barriers installed to ensure that inadvertent loading of fuel assemblies into these locations does not occur. This configuration ensures keff will remain less than or equal to 0.95 when flooded with unborated water and less than or equal to 0.98 under optimum moderation conditions. DRAINAGE 5.6.2 The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2091 fuel assemblies. In addition, no more than 225 fuel assemblies will be stored in a rack module in the cask loading area of the cask pit. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 DELETED

August 2, 2006 SEQUOYAH - UNIT 1 5-5b Amendment No. 167, 225, 309 4.3 4.3.1 4.3.2 4.3.3 4.3.1.2.a 4.3.1.2.d 4.3.1.2.b 4.3.1.2.c Page 6 of 16 A01ITS ITS Chapter 4.0 Figure 5.6-4 New Fuel Pit Storage Rack Loading Pattern

December 19, 2000 SEQUOYAH - UNIT 1 5-5g Amendment No. 225 Page 7 of 16 Figure 4.3.1.2-1 A01ITS ITS Chapter 4.0 THIS PAGE INTENTIONALLY DELETED

August 2, 2006 SEQUOYAH - UNIT 1 5-6 Amendment No. 36, 114, 157, 309 Page 8 of 32 Page 8 of 16 ITS Chapter 4.0 A01ITS 5.0 DESIGN FEATURES 5.1 SITE LOCATION The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant. EXCLUSION AREA 5.1.1 DELETED LOW POPULATION ZONE 5.1.2 DELETED SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 DELETED SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 DELETED 5.2 CONTAINMENT 5.2.1 DELETED DESIGN PRESSURE AND TEMPERATURE 5.2.2 DELETED

August 2, 2006 SEQUOYAH - UNIT 2 5-1 Amendment No. 298 4.1 Page 9 of 16 ITS Chapter 4.0 A01ITS

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August 2, 2006 SEQUOYAH - UNIT 2 5-2 Amendment No. 34, 104, 298 Page 10 of 16 ITS Chapter 4.0 A01ITS

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August 2, 2006 SEQYOYAH - UNIT 2 5-3 Amendment No. 298 Page 11 of 16 ITS Chapter 4.0 A01ITS DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Sequoyah is authorized to place a limited number of lead test assemblies into the reactor, as described in the Framatome Cogema Fuels Report BAW-2328, beginning with the Unit 2 Operating Cycle 10 core. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 DELETED VOLUME 5.4.2 DELETED 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 DELETED

August 2, 2006 SEQUOYAH - UNIT 2 5-4 Amendment No 37, 125, 172, 234, 249, 298 4.2 4.2.1 4.2.2 Page 12 of 16 ITS Chapter 4.0 A01ITS DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed for fuel enriched to 5 weight percent U-235 and shall be maintained with: a. A keff less than critical when flooded with unborated water and a keff less than or equal to 0.95 when flooded with water containing 300 ppm soluble boron.* b. A nominal 8.972 inch center-to-center distance between fuel assemblies placed in the storage racks.

c. Arrangements of one or more of three different arrays (Regions) or sub-arrays as illustrated in Figures 5.6-1 and 5.6-1a. These arrangements in the spent fuel storage pool have the following definitions:

1. Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.95 +/- 0.05 wt% U-235, (or spent fuel regardless of the fuel burnup), in a 1-in-4 checkerboard arrangement of 1 fresh assembly with 3 spent fuel assemblies with enrichment-burnup and cooling times illustrated in Figure 5.6-2 and defined by the equations in Table 5.6-1. Cooling time is defined as the period since reactor shutdown at the end of the last operating cycle for the discharged spent fuel assembly. The presence of a removable, non-fissile insert such as a burnable poison rod assembly (BPRA) or either gadolinia or integral fuel burnable absorber (IFBA) in a fresh fuel assembly does not affect the applicability of Figure 5.6-2 or Table 5.6-1. Two alternative storage arrays (or sub-arrays) are acceptable in Region 1 if the fresh fuel assemblies contain rods with either gadolinia or integral fuel burnable absorber (IFBA). For these types of assemblies, the minimum burnup of the spent fuel in the 1-of-4 sub-array are defined by the equations in Table 5.6-2. Restrictions in Region 1 Any of the three sub-arrays illustrated in Figure 5.6-1a may be used in any combination provided that: 1) Each sub-array of 4 fuel assemblies includes, in addition to the fresh fuel assembly, 3 assemblies with enrichment and minimum burnup requirements defined by the equations in Tables 5.6-1 and 5.6-2, as appropriate. 2) The arrangement of Region 1 sub-arrays must not allow a configuration with fresh assemblies adjacent to each other. 3) If Region 1 arrays are used in conjunction with Region 2 or Region 3 arrangements (see below), the arrangements shall not allow fresh fuel assemblies to be adjacent to each other (see also Figure 5.6-1). *For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident. December 19, 2000 SEQUOYAH - UNIT 2 5-5 Amendment No. 4, 52, 125, 157, 256 4.3 4.3.1 4.3.1.1 4.3.1.1.a 4.3.1.1.b 4.3.1.1.c 4.3.1.1.b See ITS 3.7.15 Page 13 of 16 ITS Chapter 4.0 A01ITS DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed for fuel enriched to 5.0 weight percent U-235 and shall be maintained with the arrangement of 146 storage locations shown in Figure 5.6-4. The cells shown as empty cells in Figure 5.6-4 shall have physical barriers installed to ensure that inadvertent loading of fuel assemblies into these locations does not occur. This configuration ensures keff will remain less than or equal to 0.95 when flooded with unborated water and less than or equal to 0.98 under optimum moderation conditions. DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2091 fuel assemblies. In addition, no more than 225 fuel assemblies will be stored in a rack module in the cask loading area of the cask pit.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 DELETED

August 2, 2006 SEQUOYAH - UNIT 2 5-5b Amendment No. 157, 216, 2984.3 4.3.1 4.3.2 4.3.3 4.3.1.2.a 4.3.1.2.d 4.3.1.2.b 4.3.1.2.c Page 14 of 16 ITS Chapter 4.0 A01ITS Figure 5.6-4 New Fuel Pit Storage Rack Loading Pattern

December 19, 2000 SEQUOYAH - UNIT 2 5-5g Amendment No. 216Page 15 of 16 Figure 4.3.1.2-1 ITS Chapter 4.0 A01ITS

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August 2, 2006 SEQUOYAH - UNIT 2 5-6 Amendment No. 28, 104, 147, 298 Page 16 of 16 DISCUSSION OF CHANGES ITS Chapter 4.0, DESIGN FEATURES Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Design Features 4.0 Westinghouse STS 4.0-1 Rev. 4.0 CTS 4SEQUOYAH UNIT 1 Amendment XXX 4.0 DESIGN FEATURES

4.1 Site Location

[ Text description of site location. ]

4.2 Reactor Core

4.2.1 Fuel Assemblies The reactor shall contain [157] fuel assemblies. Each assembly shall consist of a matrix of [Zircalloy or ZIRLO] fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 [Control Rod] Assemblies The reactor core shall contain [48] [control rod] assemblies. The control material shall be [silver indium cadmium, boron carbide, or hafnium metal] as approved by the NRC. 4.3 Fuel Storage

4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of [4.5] weight percent, b. keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR], [ c. A nominal [9.15] inch center to center distance between fuel assemblies placed in [the high density fuel storage racks], ]

    [ d. A nominal [10.95] inch center to center distance between fuel assemblies placed in [low density fuel storage racks], ]  INSERT 1 1193INSERT 2INSERT 353 full length and no part length 8.972INSERT 4 5.1 5.3 5.3.1 5.3.2 5.6 5.6 5.6.1.1 5.6.1.1 5.6.1.1.a,  Footnote

5.6.1.1.b 5.0;.1212113211M5 clad 5Zircaloy 4.0 Insert Page 4.0-1 CTS INSERT 1 The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant.

INSERT 2 Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12.

INSERT 3 The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

INSERT 4

A keff less than critical when flooded with unborated water and a keff less than or equal to 0.95 when flooded with water containing 300 ppm soluble boron. For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident; and 15.1 5.6.1.1.a, Footnote

5.3.2 5.3.1 222 Design Features 4.0 Westinghouse STS 4.0-2 Rev. 4.0 CTS 4SEQUOYAH UNIT 1 Amendment XXX 4.0 DESIGN FEATURES

4.3 Fuel Storage (continued)

    [ e. New or partially spent fuel assemblies with a discharge burnup in the "acceptable range" of Figure [3.7.17-1] may be allowed unrestricted storage in [either] fuel storage rack(s), and ]       [ f. New or partially spent fuel assemblies with a discharge burnup in the "unacceptable range" of Figure [3.7.17-1] will be stored in compliance with the NRC approved [specific document containing the analytical methods, title, date, or specific configuration or figure]. ]     4.3.1.2 The new fuel storage racks are designed and shall be maintained with:       a. Fuel assemblies having a maximum U-235 enrichment of [4.5] weight percent,

b. keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR],

c. keff 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR], and

d. A nominal [10.95] inch center to center distance between fuel assemblies placed in the storage racks.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation [23 ft]. 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than [1737] fuel assemblies. under optimum moderation conditions;INSERT 5 5.07225.6 5.6.1.2 5.6.1.2 5.6.1.2 5.6.1.2 5.6.1.2 5.6. 2 5.6.3 ;; 111313212321112091 INSERT 62INSERT 72 4.0 Insert Page 4.0-2a CTS INSERT 5 The arrangement of 146 storage locations shown in Figure 4.3.1.2-1. The cells shown as empty cells in Figure 4.3.1.2-1 shall have physical barriers installed to ensure that inadvertent loading of fuel assemblies into these locations does not occur.

INSERT 6

In addition, no more than 225 fuel assemblies will be stored in a rack module in the cask loading area of the cask pit. 5.6.1.2 225.6.3 4.0 Insert Page 4.0-2b CTS INSERT 7 Basic Cell 21 inch X 21 inch Empty Cell 9 - 4 X 5 Cell Racks 146 / 180 Loading Pattern

Figure 4.3.1.2-1 New Fuel Storage Rack Loading Pattern Figure 5.6-4 2 Design Features 4.0 Westinghouse STS 4.0-1 Rev. 4.0 CTS 4SEQUOYAH UNIT 2 Amendment XXX 4.0 DESIGN FEATURES

4.1 Site Location

[ Text description of site location. ]

4.2 Reactor Core

4.2.1 Fuel Assemblies The reactor shall contain [157] fuel assemblies. Each assembly shall consist of a matrix of [Zircalloy or ZIRLO] fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 [Control Rod] Assemblies The reactor core shall contain [48] [control rod] assemblies. The control material shall be [silver indium cadmium, boron carbide, or hafnium metal] as approved by the NRC. 4.3 Fuel Storage

4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of [4.5] weight percent, b. keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR], [ c. A nominal [9.15] inch center to center distance between fuel assemblies placed in [the high density fuel storage racks], ]

    [ d. A nominal [10.95] inch center to center distance between fuel assemblies placed in [low density fuel storage racks], ]  INSERT 1 1193INSERT 2INSERT 353 full length and no part length 8.972INSERT 4 5.1 5.3 5.3.1 5.3.2 5.6 5.6 5.6.1.1 5.6.1.1 5.6.1.1.a,  Footnote

5.6.1.1.b 5.0;.1212113211M5 clad 5Zircaloy 4.0 Insert Page 4.0-1 CTS INSERT 1 The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant. INSERT 2 Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12. INSERT 3 The full length control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. INSERT 4 A keff less than critical when flooded with unborated water and a keff less than or equal to 0.95 when flooded with water containing 300 ppm soluble boron. For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident; and 15.1 5.6.1.1.a, Footnote

5.3.2 5.3.1 222 Design Features 4.0 Westinghouse STS 4.0-2 Rev. 4.0 CTS 4SEQUOYAH UNIT 2 Amendment XXX 4.0 DESIGN FEATURES

4.3 Fuel Storage (continued)

    [ e. New or partially spent fuel assemblies with a discharge burnup in the "acceptable range" of Figure [3.7.17-1] may be allowed unrestricted storage in [either] fuel storage rack(s), and ]       [ f. New or partially spent fuel assemblies with a discharge burnup in the "unacceptable range" of Figure [3.7.17-1] will be stored in compliance with the NRC approved [specific document containing the analytical methods, title, date, or specific configuration or figure]. ]     4.3.1.2 The new fuel storage racks are designed and shall be maintained with:       a. Fuel assemblies having a maximum U-235 enrichment of [4.5] weight percent,

b. keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR],

c. keff 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR], and

d. A nominal [10.95] inch center to center distance between fuel assemblies placed in the storage racks.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation [23 ft]. 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than [1737] fuel assemblies. under optimum moderation conditions;INSERT 5 5.07225.6 5.6.1.2 5.6.1.2 5.6.1.2 5.6.1.2 5.6.1.2 5.6. 2 5.6.3 ;; 111313212321112091 INSERT 62INSERT 72 4.0 Insert Page 4.0-2a CTS INSERT 5 The arrangement of 146 storage locations shown in Figure 4.3.1.2-1. The cells shown as empty cells in Figure 4.3.1.2-1 shall have physical barriers installed to ensure that inadvertent loading of fuel assemblies into these locations does not occur.

INSERT 6

In addition, no more than 225 fuel assemblies will be stored in a rack module in the cask loading area of the cask pit. 5.6.1.2 225.6.3 4.0 Insert Page 4.0-2b CTS INSERT 7 Basic Cell 21 inch X 21 inch Empty Cell 9 - 4 X 5 Cell Racks 146 / 180 Loading Pattern

Figure 4.3.1.2-1 New Fuel Storage Rack Loading Pattern Figure 5.6-4 2 JUSTIFICATION FOR DEVIATIONS ITS Chapter 4.0, DESIGN FEATURES Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. ISTS 4.0 has been changed to address Sequoyah Nuclear Plant (SQN) site specific requirements for fuel assemblies, control rod assemblies, and fuel storage. This change is acceptable because it reflects the current licensing basis. 3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3. 4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description. 5. Typographical/grammatical error corrected.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS Chapter 4.0, DESIGN FEATURES Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ENCLOSURE 2 VOLUME 16 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision 0 LIST OF ATTACHMENTS 1. ITS Chapter 5.1 - Responsibility 2. ITS Chapter 5.2 - Organization 3. ITS Chapter 5.3 - Unit Staff Qualifications 4. ITS Chapter 5.4 - Procedures

5. ITS Chapter 5.5 - Programs and Manuals

6. ITS Chapter 5.6 - Reporting Requirements

7. ITS Chapter 5.7 - High Radiation Area ATTACHMENT 1 ITS 5.1, RESPONSIBILITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01ITS ITS 5.1 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function.

6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured. c. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 FACILITY STAFF a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units. b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. February 16, 2001 SEQUOYAH - UNIT 1 6-1 Amendment No. 32, 58, 74, 152, 178, 212, 233, 266 Page 1 of 4 See ITS 5.2 M01INSERT 1M02LA01LA02LA015.1.1 5.1.2 5.1 Insert Page 6-1 INSERT 1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. M01Page 2 of 4 A01ITS ITS 5.1 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function. 6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured. c. The Plant Manager shall be responsible for overall unit safe operation, and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 FACILITY STAFF a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units. b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. February 16, 2001 SEQUOYAH - UNIT 2 6-1 Amendment No. 24, 50, 66, 142, 169, 202, 223, 257 INSERT 1 M01LA01M02See ITS 5.2 LA02LA01Page 3 of 4 5.1.1 5.1.2 5.1 Insert Page 6-1 INSERT 1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. M01Page 4 of 4 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.1.1 states that the Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. ITS 5.1.1 states that the plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. Additionally, it requires that the plant manager or his designee approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. This changes the CTS by adding an approval requirement for the plant manager or his designee. The purpose of the ITS 5.1.1 requirement is to provide additional assurance that the plant manager has direct responsibility for overall unit operation. This change is acceptable because having the plant manager or his designee approve actions affecting nuclear safety is consistent with CTS 6.2.1.c (ITS 5.2.1.b) requirement that the plant manager shall be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. This change is designated as more restrictive because it adds a requirement for the plant manager or his designee to the CTS. M02 CTS 6.1.2 allows a designated individual to assume the responsibility for the control room command function when the Shift Manager is absent from the Control Room. ITS 5.1.2 provides the allowance for the designated individual to assume the responsibility for the control room command function, but provides additional requirements for the designated individual. In MODE 1, 2, 3, or 4, ITS 5.1.2 requires the designated individual to hold an active Senior Reactor Operator license. In MODE 5 or 6, ITS 5.1.2 requires the designated individual to hold an active Senior Reactor Operator license or Reactor Operator license. This changes the CTS by adding qualification requirements for the designated individual that assumes the control room command function. The purpose of the ITS 5.1.2 requirement is to ensure that the control room command function is maintained. This change is acceptable because the additional requirements ensure that the designated individual assuming the control room command functions meets the appropriate qualification DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES Sequoyah Unit 1 and Unit 2 Page 2 of 3 requirements. This change is designated as more restrictive because it adds qualification requirements for the designated individual that assumes the control room command function to the CTS.

RELOCATED SPECIFICATIONS None

REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.1.1 uses the title "Plant Manager" and CTS 6.1.2 uses the title "Shift Manager." ITS 5.1.1 uses the generic title "plant manager" and ITS 5.1.2 uses the generic title "shift manager." This changes the CTS by moving the specific organizational titles to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific SQN organizational titles out of the Technical Specifications is consistent with the NRC letter from C Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994. The various requirements of the plant manager and shift manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the Nuclear Power Organization Topical Report (TVA-NPOD89-A) as described in ITS 5.2.1.a. Any changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) are made under 10 CFR 50.54(a)(3), which ensures that changes are properly evaluated. This change is a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.1.3 states that the Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. ITS 5.1 does not contain this requirement. This changes the CTS by moving the requirements of the Chief Nuclear Officer to the Nuclear Power Organization Topical Report (TVA-NPOD89-A). The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the UFSAR. Changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) are made under 10 CFR 50.54(a)(3), which ensures DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES Sequoyah Unit 1 and Unit 2 Page 3 of 3 that changes are properly evaluated. This change is a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Responsibility 5.1 Westinghouse STS 5.1-1 Rev. 4.0 SEQUOYAH UNIT 1 Amendment XXX 35.0 ADMINISTRATIVE CONTROLS

5.1 Responsibility

  ---------------------------------------REVIEWER'S NOTES---------------------------------------    1. Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

2. The ANSI Standard shall be the same ANSI Standard referenced in Section 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title. Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan. ------------------------------------------------------------------------------------------------------------

5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command function. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the [SS] from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. 6.1.1 6.1.2 DOC M01 shift manager shift manager 212shift manager CTS Responsibility 5.1 Westinghouse STS 5.1-1 Rev. 4.0 SEQUOYAH UNIT 2 Amendment XXX 35.0 ADMINISTRATIVE CONTROLS

5.1 Responsibility

  ---------------------------------------REVIEWER'S NOTES---------------------------------------    1. Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

2. The ANSI Standard shall be the same ANSI Standard referenced in Section 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title. Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan. ------------------------------------------------------------------------------------------------------------

5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command function. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the [SS] from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. 6.1.1 6.1.2 DOC M01 shift manager shift manager 212shift manager CTS JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 2 ITS 5.2, ORGANIZATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS ITS 5.2 A016.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function.

6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured. c. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 FACILITY STAFF a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units. b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. February 16, 2001 SEQUOYAH - UNIT 1 6-1 Amendment No. 32, 58, 74, 152, 178, 212, 233, 266 Page 1 of 10 See ITS 5.1 5.2.1 5.2.1.a 5.2.1.c 5.2.1.b 5.2.1.d 5.2.2 5.2.2.a 5.2.2.b , respectivelyA01A01M01and established throughout, including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, A01,of the plant A01activitiesthese individualsA01The unit staff organization shall include the following:Unit or performA01LA01LA01A specified corporate officer A01 ITS ITS 5.2 A01ADMINISTRATIVE CONTROLS c. A Radiological Control technician# shall be onsite when fuel is in the reactor.

d. DELETED

e. DELETED

f. The Operations Superintendent shall hold a Senior Reactor Operator license. g. DELETED

h. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. #The Radiological Control technician may be offsite for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

February 2, 2010 SEQUOYAH - UNIT 1 6-2 Amendment No. 32, 58, 74, 152, 156, 178, 227, 233, 240, 266, 281, 327 Page 2 of 10 5.2.2,c 5.2.2.d 5.2.2.e 5.2.2.c position A01vacantnot more thanprovide for A01 ITS ITS 5.2 A01 Table 6.2-1 MINIMUM SHIFT CREW COMPOSITION WITH UNIT 2 IN MODE 5 OR 6 OR DE-FUELED

THIS PAGE INTENTIONALLY DELETED

February 16, 2001 SEQUOYAH - UNIT 1 6-3 Amendment No. 32, 58, 74, 178, 266 Page 3 of 10 ITS ITS 5.2 A01TABLE 6.2-1 (Continued) TABLE NOTATION

THIS PAGE INTENTIONALLY DELETED

February 16, 2001 SEQUOYAH - UNIT 1 6-4 Amendment No. 58, 74, 178, 266 Page 4 of 10 ITS ITS 5.2 A01ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977). 6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

February 11, 2003 SEQUOYAH - UNIT 1 6-5 Amendment No. 12, 58, 74, 119, 152, 163, 178, 212, 233, 266, 281 Page 5 of 10 See ITS 5.3 ITS ITS 5.2 A016.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function. 6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured. c. The Plant Manager shall be responsible for overall unit safe operation, and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 FACILITY STAFF a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units. b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. February 16, 2001 SEQUOYAH - UNIT 2 6-1 Amendment No. 24, 50, 66, 142, 169, 202, 223, 257 Page 6 of 10 See ITS 5.1 5.2.1 5.2.1.a 5.2.1.c 5.2.1.b 5.2.1.d 5.2.2 5.2.2.a 5.2.2.b , respectivelyA01of the plant A01activitiesthese individualsA01The unit staff organization shall include the following:Unit or performA01LA01A01 and established throughoutA01,M01 , including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, LA01A specified corporate officer ITS ITS 5.2 A01ADMINISTRATIVE CONTROLS c. A Radiological Control technician# shall be onsite when fuel is in the reactor.

d. DELETED

e. DELETED

f. The Operations Superintendent shall hold a Senior Reactor Operator license. g. DELETED

h. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. # The Radiological Control technician may be offsite for a period of time not to exceed 2 hours in 01order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

February 2, 2010 SEQUOYAH - UNIT 2 6-2 Amendment No. 50, 66, 142, 145 169, 218, 223, 230, 257, 272, 320 5.2.2,c 5.2.2.d 5.2.2.e 5.2.2.c position vacantnot more thanA01provide for A01Page 7 of 10 ITS ITS 5.2 A01TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION WITH UNIT 1 IN MODE 5 OR 6 OR DE-FUELED THIS PAGE INTENTIONALLY DELETED

February 16, 2001 SEQUOYAH - UNIT 2 6-3 Amendment No. 50, 66, 169, 257 Page 8 of 10 ITS ITS 5.2 A01TABLE 6.2-1 (Continued) TABLE NOTATION

THIS PAGE INTENTIONALLY DELETED

February 16, 2001 SEQUOYAH - UNIT 2 6-4 Amendment Nos. 50, 66, 169, 257 Page 9 of 10 ITS ITS 5.2 A01 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977).

6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m). 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 RADIOLOGICAL ASSESSMENT REVIEW COMMITTEE (RARC) (DELETED

February 11, 2003 SEQUOYAH - UNIT 2 6-5 Amendment No. 34, 50, 66, 108, 142, 153, 169, 189, 202, 223, 257, 272 See ITS 5.3 Page 10 of 10 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Sequoyah Unit 1 and Unit 2 Page 1 of 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.2.1.a regarding documentation and updating of the relationships between operating organization position, requires the organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions to be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). ITS 5.2.1.a states "These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the Nuclear Power Organization Topical Report (TVA NPOD89-A). This changes the CTS by requiring that the specific SQN organizational titles be specified in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). This change is acceptable because specifying the relationship of the specific SQN organizational titles to the generic titles used in the Technical Specifications and organizational positions, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions used in the Technical Specifications and industry standards in the Nuclear Power Organization Topical Report (TVA-NPOD89-A) continues to ensure that organizational positions and associated responsibilities will be maintained. This change adds the requirements to the Technical Specifications. This change is designated as more restrictive because it requires additional information be maintained in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.2.1.b uses the title "Chief Nuclear Officer," and CTS 6.2.1.c uses the title "Plant Manager." ITS 5.2.1.b uses the generic title DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Sequoyah Unit 1 and Unit 2 Page 2 of 2 "plant manager," and ITS 5.2.1.c uses the generic title "A specified corporate officer." This changes the CTS by moving the specific SQN organizational titles to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications, is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific SQN organizational titles out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairman, dated November 10, 1994. The various requirements of the plant manager and the specified corporate officer are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). Any changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) will be made under 10 CFR 50.54(a)(3) which will ensure the changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to meeting Technical Specification requirements is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Organization 5.2 Westinghouse STS 5.2-1 Rev. 4.0 4Amendment XXX SEQUOYAH UNIT 1 CTS 5.0 ADMINISTRATIVE CONTROLS

5.2 Organization

5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan],

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant,

c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety, and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures. 5.2.2 Unit Staff The unit staff organization shall include the following: a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4.

  ----------------------------------------REVIEWER'S NOTE----------------------------------------    Two unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units.    ------------------------------------------------------------------------------------------------------------ Nuclear Power Organization Topical Report (TVA-NPOD89-A);6.2.1 6.2.1.a 6.2.1.c 6.2.1.b 6.2.1.d 6.2.2 6.2.2.a 123;;22are required ,22unit4 Organization 5.2    Westinghouse STS 5.2-2 Rev. 4.0   4SEQUOYAH UNIT 1 Amendment XXX CTS 5.2 Organization

5.2.2 Unit Staff (continued)

b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. c. A radiation protection technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position. d. The operations manager or assistant operations manager shall hold an SRO license. e. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

6.2.2 6.2.2.b 6.2.2.c 6.2.2.f 6.2.2.h 4superintendent Specifications 5 Organization 5.2 Westinghouse STS 5.2-1 Rev. 4.0 4Amendment XXX SEQUOYAH UNIT 2 CTS 5.0 ADMINISTRATIVE CONTROLS

5.2 Organization

5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant. a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan],

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant,

c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety, and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures. 5.2.2 Unit Staff The unit staff organization shall include the following: a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4.

  ----------------------------------------REVIEWER'S NOTE----------------------------------------    Two unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units.    ------------------------------------------------------------------------------------------------------------ Nuclear Power Organization Topical Report (TVA-NPOD89-A);6.2.1 6.2.1.a 6.2.1.c 6.2.1.b 6.2.1.d 6.2.2 6.2.2.a 123;;22are required ,22unit4 Organization 5.2    Westinghouse STS 5.2-2 Rev. 4.0   4SEQUOYAH UNIT 2 Amendment XXX CTS 5.2 Organization

5.2.2 Unit Staff (continued)

b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. c. A radiation protection technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position. d. The operations manager or assistant operations manager shall hold an SRO license. e. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

6.2.2 6.2.2.b 6.2.2.c 6.2.2.f 6.2.2.h 4superintendent Specifications 5 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3. 3. ISTS 5.2.1.a contains a Reviewer's Note that allows two units with both units shutdown or defueled to have a total of three non-licensed operators for the two units. This Note applies to Sequoyah Nuclear Plant (SQN) since it is a two unit plant.

Additionally, CTS 6.2.2.a contains this same statement. Therefore, the Reviewer's Note has been deleted and the information contained in the note has been added to ITS 5.2.1.a. 4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

5. Grammatical/editorial change made for enhanced clarity.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 3 ITS 5.3, UNIT STAFF QUALIFICATIONS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS ITS 5.3 A01ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977). 6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

February 11, 2003 SEQUOYAH - UNIT 1 6-5 Amendment No. 12, 58, 74, 119, 152, 163, 178, 212, 233, 266, 281 See ITS 5.2 5.3 5.3.1 5.3.2 Page 1 of 4 ITS ITS 5.3 A01ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 THIS SPECIFICATION IS DELETED 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES & PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

May 24, 2002 SEQUOYAH - UNIT 1 6-6 Amendment No. 36, 42, 58, 74, 152, 163, 178, 198, 212, 233, 276 See ITS 5.4 Page 2 of 4 ITS ITS 5.3 A01ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977). 6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m). 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 RADIOLOGICAL ASSESSMENT REVIEW COMMITTEE (RARC) (DELETED

February 11, 2003 SEQUOYAH - UNIT 2 6-5 Amendment No. 34, 50, 66, 108, 142, 153, 169, 189, 202, 223, 257, 272 Page 3 of 4 See ITS 5.2 See ITS 5.2 5.3 5.3.1 5.3.2 ITS ITS 5.3 A01ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations. c. Surveillance and test activities of safety related equipment.

d. DELETED

e. DELETED f. Fire Protection Program implementation.

g. DELETED

May 24, 2002 SEQUOYAH - UNIT 2 6-6 Amendment No. 28, 50, 66, 142, 223, 267 See ITS 5.4 Page 4 of 4 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Unit Staff Qualifications 5.3 Westinghouse STS 5.3-1 Rev. 4.0 CTS 3SEQUOYAH UNIT 1 Amendment XXX 5.0 ADMINISTRATIVE CONTROLS

5.3 Unit Staff Qualifications

REVIEWER'S NOTE------------------------------------------------- Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures. -------------------------------------------------------------------------------------------------------------------------------

5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of [Regulatory Guide 1.8, Revision 2, 1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. [The staff not covered by Regulatory Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff]. 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m). 6.3 6.3.1 6.3.2 INSERT 112 ITS 5.3 Insert Page 5.3-1 INSERT 1 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977). 2 Unit Staff Qualifications 5.3 Westinghouse STS 5.3-1 Rev. 4.0 CTS 3SEQUOYAH UNIT 2 Amendment XXX 5.0 ADMINISTRATIVE CONTROLS

5.3 Unit Staff Qualifications

REVIEWER'S NOTE------------------------------------------------- Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures. -------------------------------------------------------------------------------------------------------------------------------

5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of [Regulatory Guide 1.8, Revision 2, 1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. [The staff not covered by Regulatory Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff]. 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m). 6.3 6.3.1 6.3.2 INSERT 112 ITS 5.3 Insert Page 5.3-1 INSERT 1 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977). 2 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis. 3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 4 ITS 5.4, PROCEDURES Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS ITS 5.4 A01ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 THIS SPECIFICATION IS DELETED 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES & PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

May 24, 2002 SEQUOYAH - UNIT 1 6-6 Amendment No. 36, 42, 58, 74, 152, 163, 178, 198, 212, 233, 276 See ITS 5.3 5.4 5.4.1 5.4.1.a Page 1 of 4 M01Add proposed Specification 5.4.1.b ITS ITS 5.4 A01ADMINISTRATIVE CONTROLS b. Refueling operations. c. Surveillance and test activities of safety-related equipment.

d. DELETED e. DELETED

f. Fire Protection Program implementation. g. DELETED

h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975.

i. OFFSITE DOSE CALCULATION MANUAL implementation.

6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The

February 11, 2003 SEQUOYAH - UNIT 1 6-7 Amendment No. 42, 58, 74, 148, 178, 233, 281 See ITS 5.5 5.4.1.d 5.4.1.c LA01M02Add proposed Specification 5.4.1.e A02A03Page 2 of 4 A02 ITS ITS 5.4 A01ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment. d. DELETED e. DELETED

f. Fire Protection Program implementation.

g. DELETED

May 24, 2002 SEQUOYAH - UNIT 2 6-6 Amendment No. 28, 50, 66, 142, 223, 267 See ITS 5.3 5.4 5.4.1 5.4.1.a M01Add proposed Specification 5.4.1.b 5.4.1.d Page 3 of 4 A02A02 ITS ITS 5.4 A01ADMINISTRATIVE CONTROLS h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975. i. OFFSITE DOSE CALCULATION MANUAL implementation. 6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following:

 (i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at lease once per 18 months.      The provisions of SR 4.0.2 are applicable

b. In-Plant Radiation Monitoring (DELETED)

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

 (i) Identification of a sampling schedule for the critical    variables and control points for these variables,      (ii) Identification of the procedures used to measure the values of    the critical variables,      (iii) Identification of process sampling points, which shall include    monitoring the discharge of the condensate pumps for evidence     of condenser in-leakage    (iv) Procedures for the recording and management of data,    (v) Procedures defining corrective actions for off-control point    chemistry conditions, (vi) Procedures identifying (a) the authority responsible for the    interpretation of the data; and (b) the sequence and timing of    administrative events required to initiate corrective action. d. Deleted February 11, 2003 SEQUOYAH - UNIT 2 6-7  Amendment No. 34, 50, 66, 134,  149, 169, 223, 272  5.4.1.c LA01A03See ITS 5.5 M02Add proposed Specification 5.4.1.e Page 4 of 4 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES  Sequoyah Unit 1 and Unit 2 Page 1 of 3 ADMINISTRATIVE CHANGES  A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 6.8.1.b requires written procedures be established, implemented and maintained covering refueling operations. CTS 6.8.1.c requires written procedures be established, implemented and maintained covering surveillance and test activities of safety-related equipment. ITS 5.4.1 requires written procedures shall be established, implemented, and maintained to the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. This changes the CTS by removing the specific wording of CTS 6.8.1.b and CTS 6.8.1.c. This change is acceptable because the recommendations of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 already require procedures for refueling operations and surveillance tests for safety related activities. This change is designated as administrative because it does not result in a technical change to the CTS. A03 CTS 6.8.1.i requires written procedures be established, implemented and maintained for the OFFSITE DOSE CALCULATION MANUAL (ODCM) implementation. ITS 5.4.1 requires procedures for various activities, but does not specifically list the ODCM. This changes the CTS by removing the specific requirement for written procedures to implement the ODCM. This change is acceptable because implementing procedures for the ODCM are required by ITS 5.4.1.e. ITS 5.4.1.e (as described in DOC M02) requires that written procedures be established, implemented and maintained for all programs and manuals listed in ITS 5.5. ITS 5.5 includes the ODCM. Therefore, it is not necessary to specifically identify each program in ITS 5.4.1. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 ITS 5.4.1.b requires that written procedures shall be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. The CTS does not include this requirement. This changes the CTS by adopting a new requirement for emergency operating procedures. DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Sequoyah Unit 1 and Unit 2 Page 2 of 3 The purpose of ITS 5.4.1.b is to ensure that written procedures are established, implemented, and maintained covering the emergency operating procedures to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. This change is acceptable because it is consistent with an existing requirement to comply with NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, for emergency operating procedures. This change is designated as more restrictive because it imposes a new requirement for procedures within the Technical Specifications. M02 ITS 5.4.1.e requires that written procedures shall be established, implemented, and maintained for all programs specified in Specification 5.5. The CTS does not include this requirement for any program except the OFFSITE DOSE CALCULATION MANUAL. This changes the CTS by adopting a new requirement for procedures to address all programs described in ITS 5.5. The purpose of ITS 5.4.1.e is to ensure that written procedures are established, implemented, and maintained covering all programs specified in ITS 5.5. This change is acceptable because it requires written procedures, including proper procedure control to address programs required by ITS 5.5. This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, or ISI Program) CTS 6.8.1.h requires written procedures be established, implemented and maintained covering the Quality Assurance Program for effluent and environmental monitoring, "using the guidance in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Revision 1, 1974, and Regulatory Guide 4.1, Revision 1, April 1975." ITS 5.4.1.c does not include the Regulatory Guide references. This changes the CTS by moving the references to the Regulatory Guides to the Nuclear Quality Assurance Program (NQAP). The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for written procedures covering quality assurance for effluent and environmental monitoring. Also, this change is acceptable because these types of procedural details will be adequately controlled in the NQAP. Any changes to the NQAP are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because references for meeting Technical Specification requirements are being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Sequoyah Unit 1 and Unit 2 Page 3 of 3 LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Procedures 5.4 Westinghouse STS 5.4-1 Rev. 4.0 4Amendment XXX SEQUOYAH UNIT 1 CTS 5.0 ADMINISTRATIVE CONTROLS

5.4 Procedures

5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in [Generic Letter 82-33], c. Quality assurance for effluent and environmental monitoring,

d. Fire Protection Program implementation, and

e. All programs specified in Specification 5.5.

6.8 6.8.1 6.8.1.a DOC M01 6.8.1.h 6.8.1.f 6.8.1.i DOC M02 ;;; ;123111program4 Procedures 5.4 Westinghouse STS 5.4-1 Rev. 4.0 4Amendment XXX SEQUOYAH UNIT 2 CTS 5.0 ADMINISTRATIVE CONTROLS

5.4 Procedures

5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in [Generic Letter 82-33], c. Quality assurance for effluent and environmental monitoring,

d. Fire Protection Program implementation, and

e. All programs specified in Specification 5.5.

6.8 6.8.1 6.8.1.a DOC M01 6.8.1.h 6.8.1.f 6.8.1.i DOC M02 ;;; ;123111program4 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3. 2. Typographical/grammatical error corrected. 3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 5 ITS 5.5, PROGRAMS AND MANUALS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ITS A01ITS 5.5 b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage). MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s). OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1. PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

February 23, 2006 SEQUOYAH - UNIT 1 1-4 Amendment No. 12, 71, 148, 155, 169, 174, 178, 281, 306 See ITS 1.0 5.5.1.a 5.5.1.b See ITS 1.0 Page 1 of 64 ITS A01ITS 5.5 SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued) If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows: Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

a. Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a; b. The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspection activities;

c. Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisions of SR 4.0.2 are not applicable); and d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirement of any TS. Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a; October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-3 Amendment No. 78, 162, 202, 208, 274, 280, 293, 301, 308 See ITS 3.0 LA015.5.6 5.5.6 5.5.6.a 5.5.5 5.5.6 LA01LA01pumps and valvesLA01pumps and valvesPage 2 of 64 A02 ITS A01ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued) b.Testing Frequencies applicable to the ASME OM Code and applicable Addenda as follows:ASME OM Code and applicable Addenda Required frequencies for terminology for inservice performing inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days c.The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normaland accelerated frequencies specified as 2 years or less in the Inservice Testing Program forperforming inservice testing activities;d.The provisions of SR 4.0.3 are applicable to inservice testing and activities; ande.Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-4 Amendment No. 78, 162, 202, 208, 274, 280, 293, 308 5.5.6.a 5.5.6.b 5.5.6.c 5.5.6.d 5.5.6 Page 3 of 64 ITS A01ITS 5.5 CONTAINMENT SYSTEMS EMERGENCY GAS TREATMENT SYSTEM - EGTS - CLEANUP SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 Two independent emergency gas treatment system cleanup subsystems (EGTS) shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one EGTS cleanup subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.8 Each EGTS cleanup subsystem shall be demonstrated OPERABLE: a.At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the controlroom, flow through the HEPA filters and charcoal adsorbers and verifying that the systemoperates for at least 10 hours with the heaters on.b.At least once per 18 months or (1) after any structural maintenance on the HEPA filter orcharcoal adsorber housings, or (2) following painting, fire or chemical release in anyventilation zone communicating with the system by:1.Verifying that the cleanup system satisfies the in-place testing acceptance criteriaand uses the test procedures of Regulatory Position C.5.a., C.5.c and C.5.d ofRegulatory Guide 1.52, Revision 2, March 1978 (except for the provisions ofANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm + 10%.2.Verifying within 31 days after removal that a laboratory analysis of arepresentative carbon sample obtained in accordance with Regulatory PositionC.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodidepenetration less than 2.5% when tested in accordance with ASTM D3803-1989 at atemperature of 30°C (86°F) and a relative humidity of 70%.3.Verifying a system flow rate of 4000 cfm + 10% during system operation whentested in accordance with ANSI N510-1975.November 2, 2000 SEQUOYAH - UNIT 1 3/4 6-13 Amendment No. 263 See ITS 3.6.10 See ITS 3.6.10 5.5.9 LA025.5.9.c 5.5.9.a 5.5.9.b Page 4 of 64 Add proposed ITS 5.5.9 generic program statementA035.5.9.a 5.5.9.b 5.5.9.c ITS A01ITS 5.5 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%. d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm +/- 10%. 2. Verifying that the filter train starts on a Phase A containment isolation Test Signal.

3. Verify the operation of the filter cooling bypass valves. 4. Verifying that each system produces a negative pressure of greater than or equal to 0.5 inches W. G. in the annulus within 1 minute after a start signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%. f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

November 2, 2000 SEQUOYAH - UNIT 1 3/4 6-14 Amendment No. 21, 88, 103, 263 See ITS 3.6.10 See ITS 3.6.7 5.5.9 LA025.5.9.c 5.5.9.d 5.5.9 5.5.9.a 5.5.9 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Page 5 of 64 ITS A01ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. 3. Verifying a system flow rate of 4000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975. d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. e. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the system at a flow rate of 4000 cfm +/- 10%. 2. Verifying that on a safety injection signal or a high radiation signal from the air intake stream, the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber banks. f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%. h. Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

October 28, 2008 SEQUOYAH - UNIT 1 3/4 7-18 Amendment No. 12, 68, 88, 263, 321 Add proposed ITS 5.5.9 generic program statement A035.5.9 LA025.5.9.c LA025.5.9 5.5.9.c 5.5.9.d See ITS 3.7.10 5.5.9 5.5.9 5.5.9.a 5.5.9.b See ITS 3.7.10 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Page 6 of 64 5.5.9 ITS A01ITS 5.5 PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated OPERABLE: a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours with the heaters on.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm +/- 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. 3. Verifying a system flow rate of 9000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 1 3/4 7-19 Amendment No. 12, 263 5.5.9 See ITS 3.7.12 LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b Page 7 of 64 Add proposed ITS 5.5.9 generic program statement A03 ITS A01ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c.After every 720 hours of charcoal adsorber operation by verifying within 31 days afterremoval that a laboratory analysis of representative carbon sample obtained in accordancewith Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows themethyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 ata temperature of 30°C (86° F) and a relative humidity of 70%.d.At least once per 18 months by:1.Verifying that the pressure drop across the combined HEPAfilters and charcoal adsorber banks is less than 3 inches Water Gauge whileoperating the filter train at a flow rate of 9000 cfm +/- 10%.2.Verifying that the filter trains start on a Containment Phase A Isolation test signal.3.Verifying that the system maintains the spent fuel storage area and the ESF pumprooms at a pressure equal to or more negative than minus 1/4 inch water gagerelative the outside atmosphere while maintaining a total system flow of 9000 cfm+/- 10%.4.Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSIN510-1975.e.After each complete or partial replacement of a HEPA filter bank by verifying that the HEPAfilter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of9000 cfm +/- 10%.f.After each complete or partial replacement of a charcoal adsorber bank by verifying that thecharcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbonrefrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 whileoperating the system at a flow rate of 9000 cfm +/- 10%.August 18, 2005 SEQUOYAH - UNIT 1 3/4 7-20 Amendment Nos. 12, 88, 103, 122, 263, 303 LA025.5.9 5.5.9.c 5.5.9.d See ITS 3.7.12 5.5.9.e 5.5.9 5.5.9.a 5.5.9 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 5.5.9 Page 8 of 64 ITS A01ITS 5.5 TABLE 4.8.1a DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) Parameter Limits for each designated pilot cell Limits for each connected cell Allowable(3) value for each connected cell Electrolyte Level >Minimum level indication mark, and 1/4" above maximum level indication mark >Minimum level indication mark, and 1/4" above maximum level indication mark Above top of plates, and not overflowing Float Voltage 2.13 volts 2.13 volts(C) > 2.07 volts 1.190 Not more than .020 below the average of all connected cells Specific Gravity(a) 1.195(b) Average of all connected cells > 1.200 Average of all connected cells > 1.190(b)

(a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery.

March 25, 1982 SEQUOYAH - UNIT 1 3/4 8-7a Amendment No. 12 Page 9 of 64 See ITS 3.8.6 See ITS 3.8.6 See ITS 3.8.6 5.5.15.b.2 LA05 ITS A01ITS 5.5 TABLE 4.8.2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) Parameter Limits for each designated pilot cell Limits for each connected cell Allowable(3) value for each connected cell Electrolyte Level >Minimum level indication mark, and 1/4" above maximum level indication mark >Minimum level indication mark, and 1/4" above maximum level indication mark Above top of plates, and not overflowing Float Voltage 2.13 volts 2.13 volts (c) > 2.07 volts 1.195 Not more than .020 below the average of all connected cells Specific Gravity(a) 1.200(b) Average of all connected cells > 1.205 Average of all connected cells > 1.195(b) (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery.

March 25, 1982 SEQUOYAH - UNIT 1 3/4 8-13a Amendment No. 12 See ITS 3.8.6 See ITS 3.8.6 Page 10 of 64 5.5.15.b.2 See ITS 3.8.6 LA05 ITS A01ITS 5.5 REFUELING OPERATIONS 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One auxiliary building gas treatment filter train shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION: a.With no auxiliary building gas treatment filter train OPERABLE, suspend all operations involvingmovement of fuel within the spent fuel pit or crane operation with loads over the spent fuel pituntil at least one auxiliary building gas treatment filter train is restored to OPERABLE status.b.The provisions of Specification 3.0.3 are not applicable.SURVEILLANCE REQUIREMENTS 4.9.12 The above required auxiliary buildings gas treatment filter train shall be demonstrated OPERABLE: a.At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room,flow through the HEPA filters and charcoal adsorbers and verifying that the system operates forat least 10 hours with the heaters on.b.At least once per 18 months or (1) after any structural maintenance on the HEPA filter orcharcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilationzone communicating with the system by:1.Verifying that the cleanup system satisfies the in-place testing acceptance criteria anduses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of RegulatoryGuide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8and 9), and the system flow rate is 9000 cfm +/- 10%.2.Verifying within 31 days after removal that a laboratory analysis of a representativecarbon sample obtained in accordance with Regulatory Position C.6.b of RegulatoryGuide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5%when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and arelative humidity of 70%.3.Verifying a system flow rate of 9000 cfm +/- 10% during system operations when tested inaccordance with ANSI N510-1975.April 11, 2005 SEQUOYAH - UNIT 1 3/4 9-12 Amendment No. 263, 301 See ITS 3.7.12 5.5.9 LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b Page 11 of 64 Add proposed ITS 5.5.9 generic program statementA03 ITS A01ITS 5.5 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) c.After every 720 hours of charcoal adsorber operation by verifying within 31 days after removalthat a laboratory analysis of representative carbon sample obtained in accordance withRegulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyliodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at atemperature of 30°C (86°F) and a relative humidity of 70%.d.At least once per 18 months by:1.Verifying that the pressure drop across the combined HEPA filters and charcoal adsorberbanks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.2.Verifying that the filter train starts on a high radiation signal from the fuel pool radiationmonitoring system.3.Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSIN510-1975.e.After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filterbanks remove greater than or equal to 99.95% of the DOP when they are tested in-place inaccordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.f.After each complete or partial replacement of a charcoal adsorber bank by verifying that thecharcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbonrefrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 whileoperating the system at a flow rate of 9000 cfm +/- 10%.November 2, 2000 SEQUOYAH - UNIT 1 3/4 9-13 Amendment No. 88, 122, 263 LA025.5.9 5.5.9.c 5.5.9.d See ITS 3.7.12 5.5.9.e 5.5.9 5.5.9.a 5.5.9 5.5.9.b Page 12 of 64 A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 5.5.9 ITS A01ITS 5.5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.1.1 This specification is deleted. 3.11.1.2 This specification is deleted. 3.11.1.3 This specification is deleted. November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-1 Amendment No. 42, 148 Page 13 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited by the following expression: i concentration of isotope i (effluent concentration limit of isotope i) excluding tritium and dissolved or entrained noble gases. a.Condensate Storage Tankb.Steam Generator Layup Tankc.Outside temporary tanks for radioactive liquidAPPLICABILITY: At all times. ACTION: a.With the quantity of radioactive material in any of the above listed tanks exceeding theabove limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. b.The provisions of Specification 3.0.3 are not applicable.SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-2 Amendment No. 42, 148, 174, 301 6,700 Add proposed ITS 5.5.10 generic program statement A04LA03A05LA03LA035.5.10 5.5.10.c 5.5.10.c The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies. A04Page 14 of 64 less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.2.1 This specification is deleted. 3.11.2.2 This specification is deleted. 3.11.2.3 This specification is deleted. 3.11.2.4 This specification is deleted. November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-3 Amendment No. 42, 65, 109, 114, 148 Page 15 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION: a.With the concentration of oxygen in a waste gas holdup tank greater than 2% by volume butless than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours. b.With the concentration of oxygen in a waste gas holdup tank greater than 4% by volumeand the hydrogen concentration greater than 2% by volume, without delay suspend all additions of waste gases to the affected waste gas holdup tank and reduce the concentration of oxygen to less than or equal to 2% by volume without delay. c.The provisions of Specification 3.0.3 are not applicable.SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by monitoring the waste gas additions to the waste gas holdup system with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-4 Amendment No. 42, 148, 301 Add proposed ITS 5.5.10 generic program statement A045.5.10.a 5.5.10 5.5.10.a LA03LA03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04Page 16 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 50,000 curies of noble gases (considered as Xe-133). APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank.

April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-5 Amendment No. 42, 148, 301 Add proposed ITS 5.5.10 generic program statementA045.5.10.b 5.5.10 5.5.10.b LA03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04LA03Page 17 of 64 to less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.3 DELETED LIMITING CONDITION FOR OPERATION 3.11.3 This specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-6 Amendment No. 42, 148 Page 18 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.4 DELETED LIMITING CONDITION FOR OPERATION 3.11.4 This specification is deleted. November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-7 Amendment No. 42, 148 Page 19 of 64 ITS A01ITS 5.5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.12.1 This Specification is deleted. 3.12.2 This Specification is deleted. 3.12.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 12-1 Amendment No. 42, 114, 148 Page 20 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS b. Refueling operations.

c. Surveillance and test activities of safety-related equipment. d. DELETED

e. DELETED f. Fire Protection Program implementation.

g. DELETED h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975. i. OFFSITE DOSE CALCULATION MANUAL implementation.

6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The

February 11, 2003 SEQUOYAH - UNIT 1 6-7 Amendment No. 42, 58, 74, 148, 178, 233, 281 5.5 5.5.2 See ITS 5.4 5.5.2 Page 21 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following: (i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at lease once per 18 months. The provisions of SR 4.0.2 are applicable. b. In-Plant Radiation Monitoring (DELETED)

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

 (i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables,         (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, (iv) Procedures for the recording and management of data,         (v) Procedures defining corrective actions for off-control point chemistry conditions,         (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action.

February 11, 2003 SEQUOYAH - UNIT 1 6-8 Amendment No. 58, 74, 178, 233, 281 5.5.2.a 5.5.2.b 5.5.2 5.5.2 5.5.8 5.5.8 5.5.8.a 5.5.8.b 5.5.8.c 5.5.8.d 5.5.8.e 5.5.8.f Page 22 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS d. DELETED e. DELETED

f. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM, 2) Limitations on the concentrations of radioactive material released in liquid effluents to, UNRESTRICTED AREAS conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5) Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

February 11, 2003 SEQUOYAH - UNIT 1 6-9 Amendment Nos. 12, 32, 58, 74, 148, 159, 174, 272, 281 5.5.3 5.5.3 5.5.3.a 5.5.3.b 5.5.3.c 5.5.3.d 5.5.3.e Page 23 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6)Limitations on the operability and use of the liquid and gaseous effluent treatmentsystems to ensure that the appropriate portions of these systems are used to reducereleases of radioactivity when the projected doses in a 31-day period would exceed 2percent of the guidelines for the annual dose or dose commitment conforming toAppendix I to 10 CFR Part 50,7)Limitations on the dose rate resulting from radioactive material released in gaseouseffluents from the site to areas at or beyond the SITE BOUNDARY shall be inaccordance with the following:1.For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the wholebody and less than or equal to a dose rate of 3000 mrem/yr to the skin, and2.For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate formwith half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.8)Limitations on the annual and quarterly air doses resulting from noble gases released ingaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming toAppendix I to 10 CFR Part 50,9)Limitations on the annual and quarterly doses to a member of the public fromIodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-livesgreater than 8 days in gaseous effluents released from each unit to areas beyond theSITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and10)Limitations on the annual dose or dose commitment to any member of the public,beyond the site boundary, due to releases of radioactivity and to radiation from uraniumfuel cycle sources conforming to 40 CFR Part 190.The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency. g.Radiological Environmental Monitoring Program (DELETED)February 11, 2003 SEQUOYAH - UNIT 1 6-10 Amendment No. 12, 32, 58, 74, 148, 174, 233, 2815.5.3.f 5.5.3.g 5.5.3.g.1 5.5.3.g.2 5.5.3.h 5.5.3.j 5.5.3.i 5.5.3 Page 24 of 64 ITS A01ITS 5.5 h.Containment Leakage Rate Testing ProgramA program shall be established to implement the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approvedexemptions. Visual examination and testing, including test intervals and extensions, shall be inaccordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-TestProgram," dated September 1995 with exceptions provided in the site implementinginstructions and the following:BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves thatare sealed with fluid from a seal system may be excluded, subject to the provisions ofAppendix J, Section III.C.3, when determining the combined leakage rate provided the sealsystem and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacityis adequate to maintain system pressure (or fluid head for the containment spray system andRHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.The peak calculated containment internal pressure for the design basis loss of coolantaccident, Pa, is 12.0 psig.The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primarycontainment air weight per day.Leakage rate acceptance criteria are: a.Containment overall leakage rate acceptance criteria is 1.0 La. During the first unitstartup following testing in accordance with this program, the leakage rate acceptancecriteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La forType A tests;b.Air lock testing acceptance criteria are:1.Overall air lock leakage rate is 0.05 La when tested at Pa.2.For each door, leakage rate is 0.01 La when pressurized to 6 psig for at least twominutes.c.For each containment purge supply and exhaust isolation valve, acceptance criteria ismeasured leakage rate less than or equal to 0.05 La.d.BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:1.The combined bypass leakage rate to the auxiliary building shall be less than orequal to 0.25 La by applicable Type B and C tests.2.Penetrations not individually testable shall have no detectable leakage when testedwith soap bubbles while the containment is pressurized to Pa (12 psig) during eachType A test.The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program. i.Configuration Risk Management Program (DELETED)April 13, 2009 SEQUOYAH - UNIT 1 6-10a Amendment No. 217, 241, 281, 287, 323 establish 5.5.14 5.5.14.a 5.5.14.a.1 5.5.14.b 5.5.14.c 5.5.14.d 5.5.14.d.1 5.5.14.d.2 1) 5.5.14.d.2 2) 5.5.14.d.2 5.5.14.f 5.5.14.e This program5.5.14.d.3 5.5.14.d.4 5.5.14.d.4 1) Page 25 of 64 5.5.14.d.4 2) 11.33 11.33 L0312.46 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of TSs. a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Programshall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). k. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met. b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage. 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full

February 23, 2006 SEQUOYAH - UNIT 1 6-11 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306 5.5.7 5.5.7.a 5.5.7.b 5.5.7.b.1 5.5.12 5.5.12 5.5.12.a 5.5.12.b 5.5.12.b.1 5.5.12.b.2 5.5.12.c 5.5.12.d 5.5.7 Page 26 of 64 ITS A01ITS 5.5 6.0 ADMINISTRATIVE CONTROLS power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. 2. Accident induced leakage performance criterion: The accident-induced leakage is not to exceed 1.0 gpm for the faulted SG. The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the maximum leakage rate established in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. 3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System, Operational Leakage." c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. 2. Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected. February 23, 2006 SEQUOYAH - UNIT 1 6-11a Amendment No. 306 5.5.7.b.2 5.5.7.b.3 5.5.7.c 5.5.7.d 5.5.7.d.1 5.5.7.d.2 5.5.7.b.1 Page 27 of 64 plugging plugging installationA06A06L01INSERT 1 ITS 5.5 Insert Page 6-11a INSERT 1 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. L01Page 28 of 64 ITS A01ITS 5.5 6.0 ADMINISTRATIVE CONTROLS 3.If crack indications are found in any SG tube, then the next inspection for each SG forthe degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. e.Provisions for monitoring operational primary-to-secondary leakage.l.Component Cyclic and Transient LimitThis program provides controls to track the FSAR, Section 5.2.1, cyclic and transientoccurrences to ensure that components are maintained within the design limits.6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4. STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS 1/ 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. 6.9.1.5 DELETED _________________ 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. August 2, 2006 SEQUOYAH - UNIT 1 6-11b Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306, 309 5.5.7.d.3 5.5.7.e 5.5.4 See ITS 5.6 Page 29 of 64 affected and potentially affectedA07results in more frequent inspectionsA07 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (DELETED) 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM: 1. Shall be documented and records of reviews performed shall be retained in a manner convenient for review. This documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. 2. Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-A.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

July 1, 1998 SEQUOYAH - UNIT 1 6-16 Amendment No. 42, 58, 74, 148, 169, 174, 178, 233 5.5.1 5.5.1 5.5.1.a 5.5.1.a.1 5.5.1.a.2 5.5.1.b 5.5.1.c plant managerM01Page 30 of 64LA04 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)** (DELETED) February 11, 2003 SEQUOYAH - UNIT 1 6-17 Amendment No. 42, 58, 74, 148, 169, 174, 233, 281 Page 31 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.16 DIESEL FUEL OIL TESTING PROGRAM A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color;

b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements: a. The definition of the CRE and the CRE boundary. b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance. c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate of 4000 cubic feet per minute plus or minus 10 percent, at a Frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

October 28, 2008 SEQUOYAH - UNIT 1 6-18 Amendment No. 261, 321 or a water and sediment content within limitsL025.5.11 5.5.11.a 5.5.11.b 5.5.11.c The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing A085.5.16 5.5.16.a 5.5.16.b 5.5.16.c 5.5.16.d required by the VFTP Page 32 of 64 A09A10LA06 ASTM D6217-11A12 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (continued) e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis. f. The provisions of SR 4.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

October 28, 2008 SEQUOYAH - UNIT 1 6-19 Amendment No. 321 Add proposed program 5.5.15 M035.5.15 5.5.16 5.5.16.e 5.5.16.f Add proposed program 5.5.17 M045.5.17 Add proposed program 5.5.13 M025.5.13 Page 33 of 64 ITS A01ITS 5.5 DEFINITIONS IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be: a. Leakage, such as that from pump seals or valve packing (except reactor coolant pump seal injection or leakoff) that is captured and conducted to collection systems or a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

May 22, 2007 SEQUOYAH - UNIT 2 1-4 Amendment Nos. 63, 134, 146, 159, 165, 169, 250, 272, 305 See ITS 1.0 5.5.1.a 5.5.1.b See ITS 1.0 Page 34 of 64 ITS A01ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued) up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows: Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

a. Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspection activities;

c. Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisions of SR 4.0.2 are not applicable); and d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirement of any TS. Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-3 Amendment No. 69, 152, 198, 263, 271, 283, 290 See ITS 3.0 5.5.6 LA01pumps and valvesLA015.5.5 LA015.5.6 5.5.6 5.5.6.a LA01pumps and valvesPage 35 of 64A02 ITS A01ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued) a.Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall beperformed in accordance with the ASME Code for Operation and Maintenance of Nuclear PowerPlants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;b.Testing frequencies applicable to the ASME OM Code and applicable Addenda as follows:ASME OMCode and applicable AddendaRequired frequencies for terminology for inserviceperforming inservice testing activitiestesting activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days c.The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normal andaccelerated frequencies specified as 2 years or less in the Inservice Test Program forperforming inservice testing activities;d.The provisions of SR 4.0.3 are applicable to inservice testing and activities; ande.Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-4 Amendment No. 69, 152, 198, 263, 271, 283, 297 5.5.6.a 5.5.6.b 5.5.6.c 5.5.6.d 5.5.6 Page 36 of 64 ITS A01ITS 5.5 CONTAINMENT SYSTEMS EMERGENCY GAS TREATMENT SYSTEM - EGTS - CLEANUP SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 Two independent emergency gas treatment system cleanup subsystems (EGTS) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one EGTS cleanup subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.8 Each EGTS cleanup subsystem shall be demonstrated OPERABLE: a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours with the heaters on. b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Position C.5.a., C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. 3. Verifying a system flow rate of 4000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 6-13 Amendment No. 254 See ITS 3.6.10 Add proposed ITS 5.5.9 generic program statementA03See ITS 3.6.10 5.5.9 LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.a 5.5.9.b 5.5.9.c Page 37 of 64 ITS A01ITS 5.5 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. d. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm + 10%. 2. Verifying that the filter train starts on a Phase A containment isolation Test Signal.

3. Verify the operation of the filter cooling bypass valves.

4. Verifying that each system produces a negative pressure of greater than or equal to 0.5 inches W.G. in the annulus within 1 minute after a start signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm + 10%. f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm + 10%.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 6-14 Amendment No. 11, 77, 92, 254 5.5.9 LA025.5.9.c 5.5.9.d See ITS 3.6.10 See ITS 3.6.7 5.5.9 5.5.9.a 5.5.9 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Page 38 of 64 ITS A01ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%.

2. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. 3. Verifying a system flow rate of 4000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ATSM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%. e. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the system at a flow rate of 4000 cfm +/- 10%.

2. Verifying that on a safety injection signal or high radiation signal from the air intake stream, the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber banks.

f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%. h. Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

October 28, 2008 SEQUOYAH - UNIT 2 3/4 7-18 Amendment No. 60, 77, 254, 313 Add proposed ITS 5.5.9 generic program statement A035.5.9 LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b LA025.5.9 5.5.9.c 5.5.9.d 5.5.9 See ITS 3.7.10 5.5.9 5.5.9 5.5.9.a 5.5.9.b See ITS 3.7.10 A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Page 39 of 64 ITS A01ITS 5.5 PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE. APPLICABILITY: Modes 1, 2, 3 and 4. ACTION: With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated OPERABLE: a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours with the heaters on. b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm + 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 9000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 7-19 Amendment No. 254 5.5.9 See ITS 3.7.12 Add proposed ITS 5.5.9 generic program statement A03LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b Page 40 of 64 ITS A01ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter trains start on a Containment Phase A Isolation test signal.

3. Verifying that the system maintains the spent fuel storage area and the ESF pump rooms at a pressure equal to or more negative than minus 1/4 inch water gauge relative the outside atmosphere while maintaining a total system flow of 9000 cfm +/- 10%.

4. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

August 18, 2005 SEQUOYAH - UNIT 2 3/4 7-20 Amendment No. 77, 111, 254, 293 LA025.5.9 5.5.9.c 5.5.9.d See ITS 3.7.12 5.5.9 5.5.9.e 5.5.9 5.5.9.a 5.5.9 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Page 41 of 64 ITS A01ITS 5.5 TABLE 4.8-1a DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS CATEROGY A(1) CATEGORY B(2) Parameter Limit for each designated pilot cell Limits for each connected cell Allowable(3) value for each connected cell Electrolyte Level >Minimum level indication mark, and 1/4" above maximum level indication mark >Minimum level indication mark, and 1/4" above maximum level indication mark Above top of plates, and not overflowing Float Voltage 2.13 volts 2.13 volts(c) > 2.07 volts 1.190 Not more than .020 below the average of all connected cells Specific Gravity(a) 1.195(b) Average of all connected cells > 1.200 Average of all connected cells 1.190(b) (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery.

SEQUOYAH - UNIT 2 3/4 8-8a Page 42 of 64 See ITS 3.8.6 See ITS 3.8.6 See ITS 3.8.6 5.5.15.b.2 LA05 ITS A01ITS 5.5 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEROGY A(1) CATEGORY B(2) Parameter Limit for each designated pilot cell Limits for each connected cell Allowable(3) value for each connected cell Electrolyte Level >Minimum level indication mark, and 1/4" above maximum level indication mark >Minimum level indication mark, and 1/4" above maximum level indication mark Above top of plates, and not overflowing Float Voltage 2.13 volts 2.13 volts(c) > 2.07 volts 1.195 Not more than .020 below the average of all connected cells Specific Gravity(a) 1.200(b) Average of all connected cells > 1.205 Average of all connected cells 1.195(b) (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery.

SEQUOYAH - UNIT 2 3/4 8-14 See ITS 3.8.6 See ITS 3.8.6 Page 39 of 58 Page 43 of 64 5.5.15.b.2 See ITS 3.8.6 LA05 ITS A01ITS 5.5 REFUELING OPERATIONS 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One auxiliary building gas treatment filter train shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION:

a. With no auxiliary building gas treatment filter train OPERABLE, suspend all operations involving movement of fuel within the spent fuel pit or crane operation with loads over the spent fuel pit until at least one auxiliary building gas treatment filter train is restored to OPERABLE status.

b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.12 The above required auxiliary building gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours with the heaters on.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm + 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 9000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 9-14 Amendment No. 254, 290 See ITS 3.7.12 5.5.9 LA025.5.9.c 5.5.9.a 5.5.9.b 5.5.9.d 5.5.9.a 5.5.9.b Add proposed ITS 5.5.9 generic program statement A03Page 44 of 64 ITS A01ITS 5.5 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%. d. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%. 2. Verifying that the filter train starts on a high radiation signal from the fuel pool radiation monitoring system.

3. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975. e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%. f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 9-15 Amendment No. 77, 111, 254 LA025.5.9 5.5.9.c 5.5.9.d See ITS 3.7.12 5.5.9.e 5.5.9 5.5.9.a 5.5.9 5.5.9.b A03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 5.5.9 Page 45 of 64 ITS A01ITS 5.5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.1.1 This Specification is deleted. 3.11.1.2 This Specification is deleted. 3.11.1.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-1 Amendment No. 34, 134 Page 46 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited by the following expression: i concentration of isotope i (effluent concentration limit of isotope i) excluding tritium and dissolved or entrained noble gases. a. Condensate Storage Tank b. Steam Generator Layup Tank

c. Outside temporary tanks for radioactive liquid APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-2 Amendment No. 134, 165, 290 6,700 I Add proposed ITS 5.5.10 generic program statement A04LA035.5.10 5.5.10.c LA03LA035.5.10.c The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies. A04Page 47 of 64A05less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.2.1 This Specification is deleted. 3.11.2.2 This Specification is deleted. 3.11.2.3 This Specification is deleted. 3.11.2.4 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-3 Amendment No. 34, 57, 99, 104, 134 Page 48 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION:

a. With the concentration of oxygen in a waste gas holdup tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.

b. With the concentration of oxygen in a waste gas holdup tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, without delay suspend all additions of waste gases to the affected waste gas holdup tank and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by monitoring the waste gas additions to the waste gas holdup system with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-4 Amendment No. 34, 134, 290 Add proposed ITS 5.5.10 generic program statement A045.5.10.a 5.5.10 5.5.10.a LA03LA03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04Page 49 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 50,000 curies of noble gases (considered as Xe-133). APPLICABILITY: At all times. ACTION:

a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-5 Amendment No. 34, 134, 290 Add proposed ITS 5.5.10 generic program statementA045.5.10.b 5.5.10 5.5.10.b LA03The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04LA03Page 50 of 64 to less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.3 DELETED LIMITING CONDITION FOR OPERATION 3.11.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-6 Amendment No. 34, 134 Page 51 of 64 ITS A01ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.4 DELETED LIMITING CONDITION FOR OPERATION 3.11.4 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-7 Amendment No. 34, 134 Page 52 of 64 ITS A01ITS 5.5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.12.1 This Specification is deleted. 3.12.2 This Specification is deleted. 3.12.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 12-1 Amendment No. 34, 104, 134 Page 53 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975. i. OFFSITE DOSE CALCULATION MANUAL implementation. 6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following:

 (i) Preventive maintenance and periodic visual inspection requirements, and    (ii) Integrated leak test requirements for each system at lease once per 18 months.      The provisions of SR 4.0.2 are applicable   b. In-Plant Radiation Monitoring (DELETED)   c. Secondary Water Chemistry    A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:    
 (i) Identification of a sampling schedule for the critical    variables and control points for these variables,

 (ii) Identification of the procedures used to measure the values of    the critical variables,

 (iii) Identification of process sampling points, which shall include    monitoring the discharge of the condensate pumps for evidence of condenser in-leakage (iv) Procedures for the recording and management of data,    (v) Procedures defining corrective actions for off-control point    chemistry conditions, (vi) Procedures identifying (a) the authority responsible for the    interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Deleted February 11, 2003 SEQUOYAH - UNIT 2 6-7 Amendment No. 34, 50, 66, 134, 149, 169, 223, 272 See ITS 5.4 5.5 5.5.2 5.5.2 5.5.2.a 5.5.2.b 5.5.2 5.5.8 5.5.8 5.5.8.a 5.5.8.b 5.5.8.c 5.5.8.d 5.5.8.e 5.5.8.f Page 54 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS e. DELETED f. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM, 2) Limitations on the concentrations of radioactive material released in liquid effluents to, UNRESTRICTED AREAS conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402,

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, 4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5) Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases

February 11, 2003 SEQUOYAH - UNIT 2 6-8 Amendment No. 28, 50, 66, 134, 165, 261, 272 5.5.3 5.5.3 5.5.3.a 5.5.3.b 5.5.3.c 5.5.3.d 5.5.3.e 5.5.3.f Page 55 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.8.4 f. Radioactive Effluent Controls Program (Cont.) of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 7)Limitations on the dose rate resulting from radioactive material released in gaseouseffluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:1.For noble gases: Less than or equal to a dose rate of 500 mrem/yr to thewhole body and less than or equal to a dose rate of 3000 mrem/yr to the skin,and2.For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate formwith half-lives greater than 8 days: Less than or equal to a dose rate of1500 mrem/year to any organ.8)Limitations on the annual and quarterly air dosesresulting from noble gasesreleased in gaseous effluents from each unit to areas beyond the SITE BOUNDARYconforming to Appendix I to 10 CFR Part 50,9)Limitations on the annual and quarterly doses to a member of the public fromIodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and10)Limitations on the annual dose or dose commitment to any member of the public,beyond the site boundary, due to releases of radioactivity and to radiation fromuranium fuel cycle sources conforming to 40 CFR Part 190.The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency.g.Radiological Environmental Monitoring Program (DELETED)h.Containment Leakage Rate Testing ProgramA program shall be established to implement the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approvedexemptions. Visual examination and testing, including test intervals and extensions, shall be inaccordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-TestProgram," dated September 1995 with exceptions provided in the site implementinginstructions and the following:BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valvesthat are sealed with fluid from a seal system may be excluded, subject to the provisions ofAppendix J, Section III.C.3, when determining the combined leakage rate provided the sealsystem and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal systemcapacity is adequate to maintain system pressure (or fluid head for the containment spraysystem and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30days.The peak calculated containment internal pressure for the design basis loss of coolantaccident, Pa, is 12.0 psig.The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primarycontainment air weight per day.April 13, 2009 SEQUOYAH - UNIT 2 6-9 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 265, 272, 276, 315 5.5.3.f 5.5.3.g 5.5.3.g.1 5.5.3.g.2 5.5.3.h 5.5.3.j 5.5.3.i 5.5.3 establish 5.5.14 5.5.14.a 5.5.14.a.1 This program5.5.14.b 5.5.14.c Page 56 of 64 11.33 12.46L03 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS Leakage rate acceptance criteria are: a.Containment overall leakage rate acceptance criteria is 1.0 La. During the first unitstartup following testing in accordance with this program, the leakage rate acceptancecriteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for TypeA tests;b.Air lock testing acceptance criteria are:1) Overall air lock leakage rate is 0.05 La when tested at Pa.2)For each door, leakage rate is 0.01 La when pressurized to 6 psig for at leasttwo minutes.c.For each containment purge supply and exhaust isolation valve, acceptance criteria ismeasured leakage rate less than or equal to 0.05 La.d.BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:1.The combined bypass leakage rate to the auxiliary building shall be less than orequal to 0.25 La by applicable Type B and C tests.2. Penetrations not individually testable shall have no detectable leakage whentested with soap bubbles while the containment is pressurized to Pa (12 psig)during each Type A test.The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program. i.Configuration Risk Management Program (DELETED)j.Technical Specification (TS) Bases Control ProgramThis program provides a means for processing changes to the Bases of these TSs.a.Changes to the Bases of the TS shall be made under appropriate administrative controlsand reviews.b.Licensees may make changes to Bases without prior NRC approval provided the changesdo not require either of the following:1.A change in the TS incorporated in the license or2.A change to the updated FSAR or Bases that requires NRC approval pursuant to10 CFR 50.59.c.The Bases Control Program shall contain provisions to ensure that the Bases aremaintained consistent with the FSAR.April 13, 2009 SEQUOYAH - UNIT 2 6-10 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 315 5.5.14.d 5.5.14.d.1 5.5.14.d.2 1) 5.5.14.d.2 2) 5.5.14.d.2 5.5.14.f 5.5.14.e 5.5.14.d.3 5.5.14.d.4 5.5.14.d.4 1) 5.5.14.d.4 2) 5.5.12 5.5.12 5.5.12.a 5.5.12.b 5.5.12.b.1 5.5.12.b.2 5.5.12.c Page 57 of 64 11.33L03 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). k. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: a. Provisions for Condition Monitoring Assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

b. Provisions for Performance Criteria for SG Tube Integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage. 1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents (DBAs). This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to- secondary pressure differential and a safety factor of 1.4 against burst applied to the DBA primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the DBAs, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The accident-induced leakage is not to exceed 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. The primary-to-secondary accident induced leakage rate for any DBA, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. 3. The operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, "Reactor Coolant System, Operational Leakage." July 10, 2012 SEQUOYAH - UNIT 2 6-10a Amendment No. 28, 50, 64, 66, 134,165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 323 5.5.12.d 5.5.7 5.5.7.a 5.5.7.b 5.5.7.b.1 5.5.7 5.5.7.b.2 5.5.7.b.3 Page 58 of 64 A11 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS c. Provisions for SG Tube Repair Criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG Tube Inspections.

Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. 2. Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SGs shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected. 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for Monitoring Operational Primary-to-Secondary Leakage. l. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

July 10, 2012 SEQUOYAH - UNIT 2 6-10b Amendment No. 28, 34, 50, 64, 66, 107, 134, 165, 207, 223, 231, 271, 272, 289, 293, 305, 315, 318, 323 5.5.7.c 5.5.7.d 5.5.7.d.1 5.5.7.d.2 5.5.7.d.3 5.5.7.e 5.5.4 Page 59 of 64 plugging plugging installationA06A06L01affected and potentially affectedA07results in more frequent inspectionsA07INSERT 1 ITS 5.5 Insert Page 6-10b INSERT 1 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. L01Page 60 of 62 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (DELETED) 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained in a manner convenient for review. This documentation shall contain: a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. 2. Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-A.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

July 1, 1998 SEQUOYAH - UNIT 2 6-17 Amendment Nos. 34, 50, 66, 134, 159, 165, 169, 223 5.5.1 5.5.1 5.5.1.a 5.5.1.a.1 5.5.1.a.2 5.5.1.b 5.5.1.c plant managerM01LA04Page 61 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)** (DELETED) February 11, 2003 SEQUOYAH - UNIT 2 6-18 Amendment Nos. 34, 50, 66, 134, 159, 165, 223, 272 Page 62 of 64 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.16 DIESEL FUEL OIL TESTING PROGRAM A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color;

b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements: a. The definition of the CRE and the CRE boundary. b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance. c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate of 4000 cubic feet per minute plus or minus 10 percent, at a Frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

October 28, 2008 SEQUOYAH - UNIT 2 6-19 Amendment No. 252, 313 or a water and sediment content within limitsL025.5.11 5.5.11.a 5.5.11.b 5.5.11.c The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing A085.5.16 5.5.16.a 5.5.16.b 5.5.16.c 5.5.16.d required by the VFTP Page 63 of 64 A09A10LA06 ASTM D6217-11A12 ITS A01ITS 5.5 ADMINISTRATIVE CONTROLS 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (continued) e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis. f. The provisions of SR 4.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

October 28, 2008 SEQUOYAH - UNIT 2 6-20 Amendment No. 313 Add proposed program 5.5.15 M035.5.15 5.5.16 5.5.16.e 5.5.16.f Add proposed program 5.5.17 M045.5.17 Add proposed program 5.5.13 M025.5.13 Page 64 of 64 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 1 of 12 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 4.0.5.c states in part, that the provisions of CTS SR 4.0.2 are not applicable for the 20 year interval reactor coolant pump flywheel inspection. ITS 5.5.5 requires a program to provide for the inspection of each reactor coolant pump flywheel. This changes the CTS by not stating that the allowance of ITS SR 3.0.2 is not applicable. This change is acceptable because no changes have been made to the existing requirements. The CTS and proposed ITS 5.5.5 continue to require the same reactor coolant pump flywheel inspections to be performed. A statement that ITS SR 3.0.2 is not applicable is not needed, as the provisions of SR 3.0.2 do not apply to the programs in ITS Section 5.5, unless specified. This change is designated as administrative because it does not result in technical changes to the CTS. A03 The Surveillances associated with the ventilation filter testing for the Control Room Ventilation System (CREVS), the Emergency Gas Treatment System (EGTS), and the Auxiliary Building Gas Treatment System (ABGTS) have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.9). As such, a general program statement has been added as ITS 5.5.9. Also, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extensions do apply (as allowed in the CTS). This changes the CTS by moving the ventilation filter testing Surveillances associated with the CREVS, EGTS, and ABGTS to a program in ITS 5.5 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. The addition of the program statement is acceptable because it is describing the intent of the CTS Surveillances. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the CTS, therefore, it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A04 The liquid holdup tank requirements in CTS 3.11.1.4, the explosive gas mixture requirements of CTS 3.11.2.5, and the gas decay tanks requirements in CTS 3.11.2.6 have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.10). As such, a general program statement has been added. Also, a statement of applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify the allowances for Surveillance Frequency extensions do apply. This changes the CTS by moving the liquid holdup tank, the explosive gas mixture, and the gas decay tanks requirements to a program in ITS 5.5.10 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 2 of 12 The addition of the program statement is acceptable because it is describing the intent of the CTS Specification. The addition of ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A05 CTS 3.11.1.4 requires that the quantity of radioactive material contained in the condensate storage tank, steam generator layup tank and outside temporary tanks for radioactive liquid shall be less than or equal to 6700 effluent concentration limit (ECL). CTS 4.11.1.4 requires a determination that the radioactive material contained in each of the tanks listed in CTS 3.11.1.4 is within limits on a prescribed frequency. ITS 5.5.10.c requires a surveillance program to ensure that the quantity of radioactive material contained in all outdoor temporary liquid radwaste storage tanks, Condensate Storage tank, and Steam Generator Layup tank is less than the amount that would result in concentrations exceeding the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. This changes the CTS by specifically stating that the program shall meet the 10 CFR 20 requirements (See DOC LA03 for discussion of the removal of the effluent concentration limit). The addition of the 10 CFR 20 limitations is acceptable because 10 CFR 20.1002 states that this part applies to persons licensed by the Commission to receive, possess, use, transfer, or dispose of byproduct, source, or special nuclear material or to operate a production or utilization facility under parts 30 through 36, 39, 40, 50, 52, 60, 61, 63, 70, or 72 of this chapter. SQN Units 1 and 2 are licensed by the Nuclear Regulatory Commission, in part, under 10 CFR Parts 30, 40, 50, and 70. 10 CFR 20.1302 requires, in part, that the annual average concentrations of radioactive material released in gaseous and liquid effluents at the boundary of the unrestricted area to not exceed the values specified in table 2 of appendix B to part 20. 10 CFR 20, Appendix B, Table 2 refers to effluent concentrations and Column 2 of this table lists limitations associated with water (liquid). Therefore, SQN Units 1 and 2 are currently required to limit effluent releases to within these concentrations. Additionally, restricting the quantity to less than or equal to 6700 ECL (See DOC LA03) provides assurance that the resulting concentrations would be less than the limits of 10 CFR 20. This change is designated as administrative because it does not result in a technical change to the CTS. A06 CTS 6.8.4.k states the requirements of the Steam Generator (SG) program. ITS 5.5.7 specifies the requirements of the Steam Generator (SG) program based on the latest revision of TSTF-510. This changes CTS 6.8.4.k.c and CTS 6.8.4.k.d by replacing the word "repair" with "plugging" and replacing the word "replacement" with "installation." CTS 6.8.4.k.d.2 has been revised to reflect TSTF-510-A. This change is acceptable because no changes have been made to the existing requirements. ITS 5.5.7 continues to require the same steam generator inspections to be performed in accordance with approved TSTF-510-A. This change is designated as administrative because it does not result in technical changes to the CTS. A07 The first sentence of CTS 6.8.4.k.d.3 states, "If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less)." The first sentence of ITS 5.5.7.d.3 states, "If crack DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 3 of 12 indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections)." The proposed change is replacing the words "for each SG" with the words "for each affected and potentially affected SG," and is replacing the parenthetical statement "(whichever is less)" with "(whichever results in more frequent inspections)". The purpose of CTS 6.8.4.k.d.3 is to restrict the allowable interval to the next scheduled inspection to 24 EFPM or one refueling outage (whichever is less) once cracks have been found in any SG tube. The intent of this requirement is that it applies to the affected SG and to any other SG which may be and affected by the degradation mechanism that caused the known crack(s). This change was made to reflect changes made under TSTF-510 and is acceptable because it clarifies the intent of the paragraph. This change is designated as administrative because it does not result in a technical change to the CTS. A08 The Diesel fuel oil testing program (CTS 6.16) has been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.11). As such, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extension do apply. This changes the CTS by specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A09 CTS 6.17.d requires, in part, that one train of the Control Room Emergency Ventilation System (CREVS) operates at a flow rate of 4000 cubic feet per minute plus or minus 10 percent. ITS 5.5.16.d requires, in part that one train of the CREVS operates at the flow rate required by the Ventilation Filter Testing Program (VFTP). This changes the CTS by requiring the CREVS to operate at the flow rate required by the VFTP. The change is acceptable because no change to the existing requirements have been made. ITS 5.5.9 contains the flow requirements for a OPERABLE CREVS train. This change is designated as administrative because it does not result in a technical change to CTS. A10 CTS 6.17.d requires, in part, measurement of the Control Room Envelope (CRE) boundary be tested using one train of the Control Room Emergency Ventilation System (CREVS) every 36 months on a STAGGERED TEST BASIS. CTS 1.35 defines a STAGGERED TEST BASIS as, "a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval." ITS 5.5.16.d requires a similar test of the CRE boundary with use of one CREVS train every 18 months "on a STAGGERED TEST BASIS." In ITS, a STAGGERED TEST BASIS consists of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 4 of 12 where n is the total number of systems, subsystems, channels, or other designated components in the associated function. This changes the CTS by utilizing the ITS definition of a STAGGERED TEST BASIS. This change is acceptable because the requirements for CRE boundary testing remain unchanged. The ITS definition of STAGGERED TEST BASIS and its application in this requirement do not change the CTS 6.17.d testing Frequency requirements. CTS 6.17.d requires each train of CREVS to be tested at least once per 36 months (one train each 18 months). ITS 5.5.16.d requires a train of CREVS to be tested each 18 months, alternating between the trains each interval. Therefore, the CTS and ITS testing Frequencies are the same. This change is designated as administrative because it does not result in technical changes to the CTS. A11 SQN Unit 2 CTS 6.8.4.k.b.2, Steam Generator (SG) Program - Accident Induced Leakage Performance Criterion, states, in part, that the accident-induced leakage is not to exceed 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. Both Unit 1 and Unit 2 CTS 6.8.4.k.b.3 contain criterion for operational leakage referencing the CTS 3.4.6.2 criterion of a maximum primary to secondary leakage of 150 gallons per day (gpd) through any one steam generator. ITS 5.5.7.b.2, Steam Generator (SG) Program - Accident Induced Leakage Performance Criterion, states, in part, that Leakage is not to exceed 1 gpm per SG while ITS 5.5.7.b.3 states that the operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE," 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). This changes the CTS by removing the duplicative non-faulted SG leakage criterion. The purpose of CTS 6.8.4.k.b is to provide provisions for performance criteria for SG tube integrity. The provisions are provided such that SG tube integrity is maintained by meeting performance criteria for tube structural integrity, accident induced leakage, and operational leakage. CTS 6.8.4.k.b.2 for Unit 1 provides one criterion for accident induced leakage, 1 gpm for the faulted SG; whereas CTS 6.8.4.k.b.2 for Unit 2 provides two criterion for accident induced leakage, 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. Both Unit 1 and Unit 2 CTS 6.8.4.k.b.3 provide an operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, "Reactor Coolant System, Operational Leakage." The CTS 3.4.6.2 requirement for the maximum primary to secondary leakage is 150 gallons per day (gpd) (0.1 gpm) through any one steam generator. The Unit 2 CTS 6.8.4.k.b.2 criterion of 0.1 gpm for each of the non-faulted SGs is duplicated in the Unit 2 CTS 6.8.4.k.b.3 criterion of the operational leakage performance criterion as reference to Limiting Condition for Operation (LCO) 3.4.6.2, "Reactor Coolant System, Operational Leakage," of 150 gpd through any one SG. This change is acceptable because duplicative leakage criterion from the Unit 2 CTS is being removed while the leakage criterion is being maintained. This change is designated as administrative because it does not result in a technical change to the CTS. A12 CTS 6.16 c requires the total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. ITS 5.5.11 c requires that the total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days but does not include a specific test method. TVA is proposing to change the test method for determining total particulate concentration for SQN to ASTM D6217-DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 5 of 12 11.This changes the CTS by requiring the testing of fuel oil total particulateconcentration to be in accordance with ASTM D6217-11. The purpose of CTS 6.16 c is to provide the requirements for testing of total particulate concentration of the fuel oil. Regulatory Guide 1.137, Revision 2, "Fuel Oil Systems for Emergency Power Supplies," describes methods that the NRC considers acceptable for use in complying with the NRC requirements regarding fuel oil systems. Based on the guidance of Regulatory Guide 1.137 ANSI/ANS-59.51 to ANSI/ASTM D2276-94 for manual sampling of the stored fuel should be changed to ASTM D6217-11. This change is acceptable because testing of total particulate concentration of the fuel oil will be done in accordance with the approved NRC method of ASTM D6217-11. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M01 CTS 6.14.1.2 states, in part, that the ODCM becomes effective after review and acceptance by the process described in TVA-NQA-PLN89-A. ITS 5.5.1.c.2 states, in part, that the ODCM becomes effective after review and acceptance by the plant manager. This changes the CTS by requiring the plant manager approval for the ODCM. The purpose of CTS 6.14.1.2 is to ensure that the ODCM has been properly reviewed by the process described in TVA-NQA-PLN89-A. ITS 5.5.1 still requires that the review process described in TVA-NQA-PLN89-A is performed (see DOC LA04 for the exclusion of the process described in TVA-NQA-PLN89-A from the ITS 5.5.1), but also includes an additional acceptance that the plant manager must review and approve the ODCM. This change is designated as more restrictive since a higher level of approval is required in the ITS than was required in the CTS. M02 The CTS does not include program requirements for the Safety Function Determination Program. The ITS includes a program for the Safety Function Determination Program. This change the CTS by adding the Safety Function Determination Program (SFDP). The Safety Function Determination Program is included to support implementation of the support system OPERABILITY characteristics of the Technical Specifications. The specific wording associated with this program is found in ITS 5.5.13. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications. M03 The CTS does not include a requirement for the Battery Monitoring and Maintenance Program. The ITS includes a requirement for this program. This changes the CTS by adding the ITS 5.5.15, "Battery Monitoring and Maintenance Program." The Battery Monitoring and Maintenance Program is included to provide for battery restoration and maintenance. The specific wording associated with this program may be found in ITS 5.5.15. This change is acceptable because it supports implementation of the requirements of the ITS. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 6 of 12 M04 The CTS does not have a Surveillance Frequency Control Program. ITS 5.5.17 requires a program to satisfy the relocation of the Surveillance Frequency from the individual specifications. This changes the CTS by incorporating the requirements of ITS 5.5.17. The NRC has been reviewing and granting improvements to the Improved Standard Technical Specifications (ISTS) based, at least in part, on probabilistic risk analysis insights. Typically, the proposed improvements involved a relaxation of one or more Completion Times or Surveillance Frequencies in the TS. In August 1995, the NRC adopted a final policy statement on the use of probabilistic risk assessment (PRA) methods, which included the following regarding the expanded use of PRA. *The use of PRA technology should be increased in all regulatory matters to the extentsupported by the state of the art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.*PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, andimportance measures) should be used in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, licensee commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule).Appropriate procedures for including PRA in the process for changing regulatoryrequirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.*PRA evaluations in support of regulatory decisions should be as realistic aspracticable and appropriate supporting data should be publicly available for review.*The Commission's safety goals for nuclear power plants and subsidiary numericalobjectives are to be used with appropriate consideration of uncertainties in makingregulatory judgments on need for proposing and backfitting new generic requirementson nuclear power plant licensees.In its approval of the policy statement, the Commission articulated its expectation that implementation of the policy statement will improve the regulatory process in three areas: foremost, through safety decision-making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees. This change is consistent with TSTF-425-A. TSTF- 425-A required that licensees who adopted this TSTF confirm that the plant PRA is consistent with Section 4.2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities." SQN has performed an assessment on the Sequoyah Units 1 and 2 PRA, and confirmed that it is consistent with the guidance in Section 4.2 of Regulatory Guide 1.200 (See 0). Future model updates (internal model or external model) will be evaluated to determine any impact on the conclusions of the assessment that was performed in support of adopting this change. For each individual Surveillance Frequency relocation, see each of the associated Technical Specifications for the DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 7 of 12 Discussion of Changes (DOC) justifying the individual relocations. This change is considered more restrictive since a new program is being added to the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 4.0.5 provides requirements for the Inservice Inspection Program. The ITS does not include Inservice Inspection Program requirements. In addition, since the Inservice Testing Program is the only requirement remaining, the reference to ASME Code Class 1, 2, and 3 "components" has been changed to "pumps and valves" for clarity. Pumps and valves are the only components related to the Inservice Testing Program (as described in CTS 4.0.5). This changes the CTS by moving these requirements from the Technical Specifications to the Inservice Inspection (ISI) Program. The removal of these requirements is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Technical Specifications still retain requirements for the affected components to be OPERABLE. Also, this change is acceptable because these requirements will be adequately controlled by the ISI, which is required by 10 CFR 50.55a. Compliance with 10 CFR 50.55a is required by the SQN Units 1 and 2 Operating Licenses. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications. LA02 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS 4.6.1.8.b.2, CTS 4.6.1.8.c, CTS 4.7.7.c.2, CTS 4.7.7.d, CTS 4.7.8.b.2, CTS 4.7.8.c, CTS 4.9.12.b.2, CTS 4.9.12.c require that within 31 days after removal of a carbon sample the laboratory analysis results are shown to be within limit. ITS 5.5.9.c requires the same analysis to be performed however the detail of "within 31 days after removal of a carbon sample" is not included. This changes the CTS by moving these procedural details from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to perform the testing at the appropriate Frequencies. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 8 of 12 LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.11.1.4 includes the details for implementing the requirements for the liquid holdup tank. CTS 3.11.2.5 includes the details for implementing the requirements for the explosive gas mixture. CTS 3.11.2.6 includes the details for implementing the requirements for the gas decay tanks. The details for implementing these requirements, including the specific limits, are not included in the ITS. CTS 3.11.2.6 Bases requires, in part in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion boundary will not exceed 0.5 rem. ITS 5.5.10.b requires the curie content in the gas decay tank to be less than the amount that would result in whole body exposure of greater than or equal to 0.5 rem at the exclusion boundary. This changes the CTS by moving the boundary exposure limit of 0.5 rem from the Bases to ITS 5.5.10 and moving those procedural details for implementing the requirements, including the specific limits, from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details for the specific limits, Applicability, Actions, and Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.10 still retains the requirement to include a program, which provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in condensate storage tank, steam generator layup tank and outdoor temporary liquid radwaste storage tanks. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA04 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, ISI Program) CTS 6.14.1.2 requires changes to the ODCM to be effective after review and acceptance by the process described in TVA-NQA-PLN89-A. ITS 5.5.1.b requires changes to the ODCM to become effective after the approval of the plant manager. This changes the CTS by moving the process described in TVA-NQA-PLN89-A to the Nuclear Facility Quality Assurance Program Description (NFQAPD). DOC M01 describes the addition of the plant manager approval. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable because these types of procedural details will be adequately controlled in the NFQAPD. Any changes to the NFQAPD are made under 10 CFR 50.54(a), which ensure changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specifications requirements are being removed from the Technical Specifications. LA05 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS Table 4.8.1a and Table 4.8.2 Unit 1 footnote (c) and CTS Table 4.8-1a and Table 4.8-2 Unit 2 footnote (c) states, in part the float voltage of 2.13 volts DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 9 of 12 is corrected for average electrolyte temperature. ITS 5.5.15 b.1 requires a program with actions to restore battery cells with float voltage < 2.13 V and ITS 5.5.15 b.2 requires a program with actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cells has been found to be < 2.13 V. This changes the CTS by by moving information from the specification to the Battery Monitoring and Maintenance Program implementing document. The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.15 still retains the requirement for float voltage 2.13 V. Also, this change is acceptable because these types of procedural details will be adequately controlled by the requirements of a program required by ITS Chapter 5. ITS 5.5.15, Battery Monitoring and Maintenance program is controlled by Chapter 5 of the Technical Specifications. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA06 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 6.16 c requires total particulate of the fuel oil to be less than or equal to 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. ITS 5.5.11.c requires the total particulate concentration of the fuel oil is less than or equal to 10 mg/l when tested every 31 days. This changes the CTS by moving the details of using of particulate testing standard ASTM D-2276, Method A from the CTS to the ITS SR 3.8.3.3 Bases. The removal of these details related to testing standards from the Technical Specification is acceptable, because this type of information is not necessary to be included in the Technical Specification to provide adequate protection of the public health and safety. The ITS retains the requirement for fuel oil particulate testing every 31 days. Also, this change is acceptable because the removed details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to the Bases to ensure the Bases are properly controlled. This change is designated as less restrictive removal of detail change, because information relating to testing standards is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 7 - Relaxation of Surveillance Frequency) CTS 6.8.4.k.d.2 states, "Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected." ITS 5.5.7.d.2 states, "After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 10 of 12 tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage." ITS 5.5.7.d.2 goes on to describe the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d by stating, "a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods." This changes the CTS by relaxing the surveillance frequency for inspecting SG tubes. The purpose of CTS 6.8.4.k is to ensure that SG tube integrity is maintained by providing provisions regarding the scope, frequency, and methods of SG tube inspections. These changes to when inspections are performed are considered marginal increases for consistency with typical fuel cycle lengths that better accommodate the scheduling of inspections and reflect the improved resistance of alloy 690 TT SG tubes to stress corrosion cracking. This change is acceptable because TVA has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," (ADAMS Accession No. ML110610350) and the model safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as part of the Federal Register Notice for Comment. As described in the subsequent paragraphs, TVA has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the NRC staff are applicable to SQN Unit 1 and Unit 2 and justify the incorporation of the changes to the SQN Unit 1 and SQN Unit 2 ITS. TVA is proposing the following variations from the TS changes described in the TSTF-510, Revision 2, or the applicable parts of the NRC staff's model safety evaluation dated October 19, 2011. SQN Unit 1 and Unit 2 ITS utilize different numbering than the Standard Technical Specifications (NUREG 1431, Revision 4.0, "Standard Technical Specifications Westinghouse Plants") on which TSTF-510 was based. The specific numbering differences are: 1) TSTF-510 Rev. 2, TS 3.4.20, "Steam Generator Tube Integrity," is ITS 3.4.17, "Steam Generator Tube Integrity"; 2) TSTF-510 Rev. 2 TS 5.5.9, "Steam Generator (SG) Program is ITS 5.5.7, "Steam Generator (SG) Program"; and 3) TSTF-510 Rev. 2 TS 5.6.7, "Steam Generator Tube Inspection Report," is ITS 5.6.6, "Steam Generator Tube Inspection Report." This change is designated as less restrictive because the SG tube inspections will be performed less frequently in ITS than they were in CTS. DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 11 of 12 L02 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 6.16.a.3 requires performance of the "clear and bright" test, used to establish the acceptability of new fuel oil for use prior to addition to storage tanks. ITS 5.5.11.a.3 requires a determination that the fuel oil has a clear and bright appearance with proper color or that water and sediment content is within limits. This changes the CTS by allowing a "water and sediment content" test to be performed to establish the acceptability of new fuel oil instead of only allowing a "clear and bright" test. CTS 6.16.a.3 requires performance of the "clear and bright" test, to establish the acceptability of new fuel oil for use prior to addition to storage tanks. ITS 5.5.11.a.3 is proposed to be expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil instead of the "clear and bright" test. ASTM D4176-86,"Standard Test Method for Free Water and Particulate Contamination in Distillate Fuels (Clear and Bright Pass/Fail Procedures)," verifies that the new fuel oil has a clear and bright appearance with proper color. The "clear and bright" test is only applicable to fuel oils that meet the ASTM D4176 color rating requirements (i.e., an ASTM D1500, "Test Method for ASTM Color of Petroleum Products (ASTM Color Scale)," color rating of five or less). The "clear and bright" test is a qualitative test for determining free water and particulate contamination in distillate fuels and is, therefore, subject to human interpretation. For example, if an attempt is made to use the qualitative "clear and bright" test with darker colored fuels (e.g., for high sulfur fuel oil that has been dyed in accordance with EPA mandated requirements), the presence of free water or particulate could be obscured and missed by the viewer. Therefore, ITS 5.5.11.a.3 has been expanded to allow a water and sediment content test. The water and sediment content test is a quantitative test using centrifuge methods. In ASTM D975-90, ASTM D1796, "Standard Method for Water and Sediment in Fuel Oils by the Centrifuge Method (Laboratory Procedure)," is an acceptable standard for the water and sediment content test. In addition, the use of ASTM D1796-83 was endorsed by the NRC in Amendment No. 101 for the Wolf Creek Generating Station. ASTM D1796-83 is the same ASTM Standard used to verify the water and sediment content is within limits within 31 days following sampling and addition to the storage tanks as required by CTS 6.16.b. Therefore, since ASTM D1796 is currently used to verify the acceptability of new fuel oil for use after addition to the storage tanks, the use of these quantitative methods (i.e., water and sediment content) in lieu of ASTM D4176 (i.e., "clear and bright" test) does not introduce a different method for determining the acceptability of new fuel oil. This change is designated as less restrictive because Surveillance acceptance criteria required in the CTS will have alternative acceptance criteria allowed in ITS. L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 6.8.4.h specifies the limit for peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig. Also, CTS 6.8.4.h specifies, in part, that bypass leakage paths to the auxiliary building from the isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig). CTS 6.8.4.h.d.2 requires penetrations not individually testable to have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test. ITS 5.5.14.a specifies, in part, that bypass leakage paths to the auxiliary building from the isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, Section III.C.3, when determining the combined DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 12 of 12 leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig). ITS 5.5.14.b specifies the calculated peak containment internal pressure for the design basis loss of coolant accident, Pa is 11.33 psig and the containment design pressure is 12.0 psig. ITS 5.5.14.d.4.b requires penetrations not individually testable to have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type A test. This changes the CTS by reducing the calculated peak containment internal pressure for the design basis loss of coolant accident, Pa to 11.33 psig and specifying the containment design pressure is 12.0 psig. The purpose of ITS 5.5.14 is to ensure the appropriate limits are specified for the Containment Leakage Rate Testing Program. This change is acceptable because the acceptable limits continue to ensure the containment leakage is within the value assumed in the accident analysis as described in the recent application to modify Ice Condenser Technical Specifications (ML13199A281). Currently, SQN is using the containment design pressure value of 12.0 psig as Pa. In the ITS, the value of Pa (11.33 psig) is the calculated peak containment internal pressure for the design basis loss of coolant accident. This is acceptable because the value of Pa (11.33 psig) is the value assumed in the accident analyses. This change is designated as less restrictive because a lower pressure will be used for Pa in the Containment Leakage Rate Testing Program. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Programs and Manuals 5.5 Westinghouse STS 5.5-1 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

The following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program, and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.1] and Specification [5.6.2].

Licensee initiated changes to the ODCM: a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s) and

2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations,

b. Shall become effective after the approval of the plant manager, and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented. 1.18 1.18 1.18 6.14.1.1 6.14.1.1.a 6.14.1.1.b 6.14.1.1 6.14.1.2 6.14.1.3 2; 1 Programs and Manuals 5.5 Westinghouse STS 5.5-2 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals

5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include [Recirculation Spray, Safety Injection, Chemical and Volume Control, gas stripper, and Hydrogen Recombiner]. The program shall include the following: a. Preventive maintenance and periodic visual inspection requirements and

b. Integrated leak test requirements for each system at least once per [18] months. The provisions of SR 3.0.2 are applicable.

5.5.3 [ Post Accident Sampling

----------------------------------------REVIEWER'S NOTE---------------------------------------- This program may be eliminated based on the implementation of WCAP-14986, Rev. 1, "Post Accident Sampling System Requirements: A Technical Basis," and the associated NRC Safety Evaluation dated June 14, 2000, and implementation of the following commitments: 1. [Licensee] has developed contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere. The contingency plans will be contained in emergency plan implementing procedures and implemented with the implementation of the License amendment. Establishment of contingency plans is considered a regulatory commitment. 2. The capability for classifying fuel damage events at the Alert level threshold has been established for [Plant] at radioactivity levels of 300 mCi/cc dose equivalent iodine. This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations. This capability will be described in emergency plan implementing procedures and implemented with the implementation of the License amendment. The capability for classifying fuel damage events is considered a regulatory commitment. 3. [Licensee] has established the capability to monitor radioactive iodines that have been released to offsite environs. This capability is described in our emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment. ------------------------------------------------------------------------------------------------------------

This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents 6.8.4.a 6.8.4.a 6.8.4.b 2Residual Heat Removal, Containment Spray, and RCS Sampling 4 Programs and Manuals 5.5 Westinghouse STS 5.5-3 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals

5.5.3 Post Accident Sampling (continued) and containment atmosphere samples under accident conditions. The program shall include the following: a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analysis equipment. ]

5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I,

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, 3 6.8.4.f 6.8.4.f 6.8.4.f.1) 6.8.4.f.2) 6.8.4.f.3) 6.8.4.f.4) 6.8.4.f.5) 6.8.4.f.6) 44 Programs and Manuals 5.5 Westinghouse STS5.5-4Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued) g.Limitations on the dose rate resulting from radioactive material released ingaseous effluents from the site to areas at or beyond the site boundary shallbe in accordance with the following:1.For noble gases: a dose rate 500 mrem/yr to the whole body and adose rate 3000 mrem/yr to the skin and2.For iodine-131, iodine-133, tritium, and all radionuclides in particulateform with half-lives greater than 8 days: a dose rate 1500 mrem/yr toany organ,h.Limitations on the annual and quarterly air doses resulting from noble gasesreleased in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I,i.Limitations on the annual and quarterly doses to a member of the publicfrom iodine-131, iodine-133, tritium, and all radionuclides in particulate formwith half lives > 8 days in gaseous effluents released from each unit toareas beyond the site boundary, conforming to 10 CFR 50, Appendix I, andj.Limitations on the annual dose or dose commitment to any member of thepublic, beyond the site boundary, due to releases of radioactivity and toradiation from uranium fuel cycle sources, conforming to 40 CFR 190.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. 5.5.5 Component Cyclic or Transient Limit This program provides controls to track the FSAR, Section [ ], cyclic and transient occurrences to ensure that components are maintained within the design limits. 5.5.6 [ Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC. The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ] 3 4 6.8.4.f.7) 6.8.4.f.7)).1 6.8.4.f.7)).2 6.8.4.f.8) 6.8.4.f.9) 6.8.4.f.10) 6.8.4.f 6.8.4.f 6.8.4.l 5.2.1 4244U Programs and Manuals 5.5 Westinghouse STS5.5-5Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals. ---------------------------------------REVIEWER'S NOTE---------------------------------------- The inspection interval and scope for RCP flywheels stated above can be applied to plants that satisfy the requirements in WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination." ------------------------------------------------------------------------------------------------------------ 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following: a.Testing frequencies applicable to the ASME Code for Operations andMaintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Required Frequencies for performing inservice testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days b.The provisions of SR 3.0.2 are applicable to the above requiredFrequencies and to other normal and accelerated Frequencies specified as2 years or less in the Inservice Testing Program for performing inservicetesting activities,c.The provisions of SR 3.0.3 are applicable to inservice testing activities, and6 5 4.0.5.c 4.0.5 4.0.5 4.0.5.b 4.0.5.c 4.0.5.d 544pumps and valves 6ultrasonic liquid penetrantmagnetic particle 33 Programs and Manuals 5.5 Westinghouse STS5.5-6Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.8 Inservice Testing Program (continued) d.Nothing in the ASME OM Code shall be construed to supersede therequirements of any TS.5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: a.Provisions for condition monitoring assessments. Condition monitoringassessment means an evaluation of the "as found" condition of the tubingwith respect to the performance criteria for structural integrity and accidentinduced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging [or repair] oftubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, [or repaired] toconfirm that the performance criteria are being met.b.Performance criteria for SG tube integrity. SG tube integrity shall bemaintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1.Structural integrity performance criterion: All in-service steamgenerator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transientsincluded in the design specification) and design basis accidents. Thisincludes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated withthe design basis accidents, or combination of accidents in accordancewith the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2on the combined primary loads and 1.0 on axial secondary loads.2.Accident induced leakage performance criterion: The primary tosecondary accident induced leakage rate for any design basisaccident, other than a SG tube rupture, shall not exceed the leakagerate assumed in the accident analysis in terms of total leakage rate forall SGs and leakage rate for an individual SG. Leakage is not to), , 6.8.4.k 6.8.4.k.a 6.8.4.k.b 6.8.4.k.b.1 6.8.4.k.b.2 22TSTF-510-A6 7 4.0.5.e 4.0.5 44or 1 Programs and Manuals 5.5 Westinghouse STS5.5-7Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) exceed [1 gpm] per SG [, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program. 3.The operational LEAKAGE performance criterion is specified inLCO 3.4.13, "RCS Operational LEAKAGE."c.Provisions for SG tube repair criteria. Tubes found by inservice inspectionto contain flaws with a depth equal to or exceeding [40%] of the nominaltube wall thickness shall be plugged [or repaired].---------------------------------------REVIEWER'S NOTE---------------------------------------- Alternate tube repair criteria currently permitted by plant technical specifications are listed here. The description of these alternate tube repair criteria should be equivalent to the descriptions in current technical specifications and should also include any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria. --------------------------------------------------------------------------------------------------------------- [The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria: 1.. . .]d.Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methodsof inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may bepresent along the length of the tube, from the tube-to-tubesheet weld at thetube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is notpart of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervalsshall be such as to ensure that SG tube integrity is maintained until the nextSG inspection. An assessment of degradation shall be performed todetermine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.6.8.4.k.b.3 6.8.4.k.c 6.8.4.k.d ] plugging [or ]] plugging [or ]]plugging [or plugging [or ]plugging [or ] assessment 22TSTF-510-ATSTF-510-A5TSTF-510-ATSTF-510-A27 4TSTF-510-A7 Programs and Manuals 5.5 Westinghouse STS 5.5-8 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued) ---------------------------------------REVIEWER'S NOTE---------------------------------------- Plants are to include the appropriate Frequency (e.g., select the appropriate Item 2.) for their SG design. The first Item 2 is applicable to SGs with Alloy 600 mill annealed tubing. The second Item 2 is applicable to SGs with Alloy 600 thermally treated tubing. The third Item 2 is applicable to SGs with Alloy 690 thermally treated tubing. ------------------------------------------------------------------------------------------------------------

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. [2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.]

   [2. Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.]      [2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.]

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary to secondary LEAKAGE. INSERT 3 INSERT 1 INSERT 2 installationaffected and potentially affected results in more frequent inspections 6.8.4.k.d.1 6.8.4.k.d.2 6.8.4.k.d.3 6.8.4.k.e 2TSTF-510-ATSTF-510-A7 54 5.5 Insert Page 5.5-8 CTS INSERT 1 After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

INSERT 2

After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

Reviewer's Note ------------------------------------------------------

A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods. ------------------------------------------------------------------------------------------------------------------------------- a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; 2TSTF-510-A2 5.5 Insert Page 5.5-8 CTS b)During the next 96 effective full power months, inspect 100% of the tubes. This constitutesthe second inspection period; and c)During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full powermonths. This constitutes the third and subsequent inspection periods. INSERT 3 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. ----------------------------------------------- Reviewer's Note ------------------------------------------------------ A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods. ----------------------------------------------------------------------------------------------------------------------------- a)After the first refueling outage following SG installation, inspect 100% of the tubes during thenext 144 effective full power months. This constitutes the first inspection period;b)During the next 120 effective full power months, inspect 100% of the tubes. This constitutesthe second inspection period;c)During the next 96 effective full power months, inspect 100% of the tubes. This constitutesthe third inspection period; andd)During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full powermonths. This constitutes the fourth and subsequent inspection periods.2TSTF-510-A Programs and Manuals 5.5 Westinghouse STS5.5-9Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) [f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below. ---------------------------------------REVIEWER'S NOTE---------------------------------------- Tube repair methods currently permitted by plant technical specifications are to be listed here. The description of these tube repair methods should be equivalent to the descriptions in current technical specifications. If there are no approved tube repair methods, this section should not be used. ------------------------------------------------------------------------------------------------------------ 1.. . .]5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include: a.Identification of a sampling schedule for the critical variables and controlpoints for these variables,b.Identification of the procedures used to measure the values of the criticalvariables,c.Identification of process sampling points, which shall include monitoring thedischarge of the condensate pumps for evidence of condenser in leakage, d.Procedures for the recording and management of data,e.Procedures defining corrective actions for all off control point chemistryconditions, andf.A procedure identifying the authority responsible for the interpretation of thedata and the sequence and timing of administrative events, which isrequired to initiate corrective action.27 8 6.8.4.c 6.8.4.c 6.8.4.c.(i) 6.8.4.c.(ii) 6.8.4.c.(iii) 6.8.4.c.(iv) 6.8.4.c.(vi) 6.8.4.c.(v) 48452; ; ; ; ; 11 Programs and Manuals 5.5 Westinghouse STS5.5-10Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, Revision 2, ASME N510-1989, and AG-1]. a.Demonstrate for each of the ESF systems that an inplace test of the highefficiency particulate air (HEPA) filters shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%]. ESF Ventilation System Flowrate [ ] [ ] b.Demonstrate for each of the ESF systems that an inplace test of thecharcoal adsorber shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%]. ESF Ventilation System Flowrate [ ] [ ] c.Demonstrate for each of the ESF systems that a laboratory test of a sampleof the charcoal adsorber, when obtained as described in [Regulatory Guide 1.52, Revision 2], shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity specified below. ESF Ventilation System Penetration RH Face Velocity (fps) [ ] [See Reviewer's [See [See Reviewer's Note] Reviewer's Note] Note] ----------------------------------------REVIEWER'S NOTE---------------------------------------- The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate. 9 INSERT 5 ANSI N510-1975 (except for the provisions of Sections 8 and 9)ANSI N510-1975Regulatory Positions C.5.a, C.5.c, C.5.d and C.6.b of ASTM D3803-19894.6.1.8.e 4.7.7.c.3 4.7.7.f 4.7.8.b.3 4.7.8.e 4.9.12.b.3 4.9.12.e removal efficiently of 99.95% of a halogenated hydrocarbon refrigerant test gas removal efficiently of 99.95% of dioctyl phthalate (DOP) ANSI N510-1975 (except for the provisions of Sections 8 and 9)4.6.1.8.f 4.7.7.c.3 4.7.7.g 4.7.8.b.3 4.7.8.f 4.9.12.b.3 4.9.12.f INSERT 6 Regulatory Position C.6.b of of 70% < 2.5%4.6.1.8.b.2 4.6.1.8.c 4.7.7.c..2 4.7.7.d 4.7.8.b.2 4.7.8.c 4.9.12.b.2 4.9.12.c EGTS ABGTS CREVS Regulatory Positions C.5.a and C.5.c of Regulatory Positions C.5.a and C.5.d of4.6.1.8.b.1 4.7.7.c.1 4.7.8.b.1 4.9.12.b.1 4323322232323323325INSERT 4 3 5.5 Insert Page 5.5-10 CTS INSERT 4 The test described in Specification 5.5.9.a and 5.5.9.b shall be performed once per 18 months; after any structural maintenance on the high efficiency particulate air (HEPA) filter bank or charcoal adsorber bank housing; following painting, fire, or chemical release in any ventilation zone communicating with the system; and after each complete or partial replacement of a HEPA filter bank or charcoal adsorber bank. The test described in Specification 5.5.9.c shall be performed once per 18 months or after 720 hours of filter operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and following painting, fire, or chemical release in any ventilation zone communicating with the system. The test described in Specification 5.5.9.d and 5.5.9.e shall be performed once per 18 months. INSERT 5 ESF Ventilation System Flow Rate (cfm) Emergency Gas Treatment System (EGTS) 4000 Auxiliary Building Gas Treatment System (ABGTS) 9000 Control Room Emergency Ventilation System (CREVS) 4000 INSERT 6 ESF Ventilation System Flow Rate (cfm) EGTS 4000 ABGTS 9000 CREVS 4000 3334.6.1.8.b 4.6.1.8.e 4.6.1.8.f 4.7.7.f 4.7.7.g 4.7.8.e 4.7.8.f 4.6.1.8.c 4.6.1.8.b 4.7.7.c 4.7.7.d 4.7.8.b 4.7.8.c 4.9.12.b 4.9.12.c 4.6.1.8.d 4.7.7.e 4.7.8.d 4.9.12.d Programs and Manuals 5.5 Westinghouse STS5.5-11Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued) ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95% (or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test. Allowable Penetration = [(100% - Methyl Iodide Efficiency

for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor] When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following: Safety factor 2 for systems with or without humidity control. Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions. If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified. *This value should be the efficiency that was incorporated in the licensee'saccident analysis which was reviewed and approved by the staff in a safety evaluation. ------------------------------------------------------------------------------------------------------------ d.Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters, the prefilters, and the charcoal adsorbers is lessthan the value specified below when tested in accordance with [RegulatoryGuide 1.52, Revision 2, and ASME N510-1989] at the system flowratespecified below [+/- 10%].ESF Ventilation System Delta P Flowrate [ ] [ ] [ ] [ e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below [+/- 10%] when tested in accordance with [ASME N510-1989]. ESF Ventilation System Wattage ] [ ] [ ] The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 9 INSERT 7 4.6.1.8.d.1 4.7.7.e.1 4.7.8.b.3 4.7.8.d.1 4.9.12.b.3 4.9.12.d.1 4.7.7.c.3 ANSI N510-1975 Auxiliary Building Gas Treatment System32 kW4.7.8.d.4 4.9.12.d.3 5323241975 5.5 Insert Page 5.5-11 CTS INSERT 7 ESF Ventilation System Combined Delta P (inches water gauge) Flowrate (cfm) EGTS54000ABGTS39000CREVS340003 Programs and Manuals 5.5 Westinghouse STS5.5-12Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous radioactivity quantities shall be determined following the methodology in [Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be determined in accordance with [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"]. The program shall include: a.The limits for concentrations of hydrogen and oxygen in the [Waste GasHoldup System] and a surveillance program to ensure the limits aremaintained. Such limits shall be appropriate to the system's design criteria(i.e., whether or not the system is designed to withstand a hydrogen explosion),b.A surveillance program to ensure that the quantity of radioactivity containedin [each gas storage tank and fed into the offgas treatment system] is lessthan the amount that would result in a whole body exposure of 0.5 rem toany individual in an unrestricted area, in the event of [an uncontrolledrelease of the tanks' contents], andc.A surveillance program to ensure that the quantity of radioactivity containedin all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the [Liquid Radwaste Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. 10 3.11.1.4 3.11.2.5 3.11.2.6 the Condensate Storage Tank, Steam Generator Layup Tank, and 4.11.2.5 4.11.2.6 4.11.1.4 storage temporary temporary29422decay decay , Condensate Storage Tank, and Steam Generator Layup Tank 32exceeding ; 11; 11 Programs and Manuals 5.5 Westinghouse STS5.5-13Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following: a.Acceptability of new fuel oil for use prior to addition to storage tanks bydetermining that the fuel oil has:1.An API gravity or an absolute specific gravity within limits,2.A flash point and kinematic viscosity within limits for ASTM 2D fuel oil,and3.A clear and bright appearance with proper color or a water andsediment content within limits.b.Within 31 days following addition of the new fuel oil to storage tanks, verifythat the properties of the new fuel oil, other than those addressed in a.,above, are within limits for ASTM 2D fuel oil, andc.Total particulate concentration of the fuel oil is 10 mg/l when tested every31 days.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies. 5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications. a.Changes to the Bases of the TS shall be made under appropriateadministrative controls and reviews.b.Licensees may make changes to Bases without prior NRC approvalprovided the changes do not require either of the following:1.A change in the TS incorporated in the license or2.A change to the updated FSAR or Bases that requires NRC approvalpursuant to 10 CFR 50.59.6.16 1112 6.16 6.16.a 6.16.a.1 6.16.a.2 6.16.a.3 6.16.b 6.16.c 6.8.4.j 6.8.4.j 6.8.4.j.a 6.8.4.j.b 6.8.4.j.b.1 6.8.4.j.b.2 44 Programs and Manuals 5.5 Westinghouse STS5.5-14Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.14 Technical Specifications (TS) Bases Control Program (continued) c.The Bases Control Program shall contain provisions to ensure that theBases are maintained consistent with the FSAR.d.Proposed changes that meet the criteria of Specification 5.5.14b above shallbe reviewed and approved by the NRC prior to implementation. Changes tothe Bases implemented without prior NRC approval shall be provided to theNRC on a frequency consistent with 10 CFR 50.71(e).5.5.15 Safety Function Determination Program (SFDP) This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following: a.Provisions for cross train checks to ensure a loss of the capability to performthe safety function assumed in the accident analysis does not go undetected,b.Provisions for ensuring the plant is maintained in a safe condition if a loss offunction condition exists, c.Provisions to ensure that an inoperable supported system's CompletionTime is not inappropriately extended as a result of multiple support system inoperabilities, and d.Other appropriate limitations and remedial or compensatory actions.A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: a.A required system redundant to the system(s) supported by the inoperablesupport system is also inoperable, or b.A required system redundant to the system(s) in turn supported by theinoperable supported system is also inoperable, or12 13 6.8.4.j.d 6.8.4.j.d 6.8.4.j.d DOC M02 4412 4U3; 11; ; ; ; ;

Programs and Manuals 5.5 Westinghouse STS5.5-15Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued) c.A required system redundant to the support system(s) for the supportedsystems (a) and (b) above is also inoperable.The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system. 5.5. 16 Containment Leakage Rate Testing Program [OPTION A] a.A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. b.The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% ofcontainment air weight per day. c.Leakage rate acceptance criteria are:1.Containment leakage rate acceptance criterion is 1.0 La. During thefirst unit startup following testing in accordance with this program, theleakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests. 2.Air lock testing acceptance criteria are:a)Overall air lock leakage rate is [0.05 La] when tested at Pa.b)For each door, leakage rate is [0.01 La] when pressurized to[ 10 psig]. d.The provisions of SR 3.0.3 are applicable to the Containment Leakage RateTesting Program. e.Nothing in these Technical Specifications shall be construed to modify thetesting Frequencies required by 10 CFR 50, Appendix J. 13 14 DOC M02 4410 Programs and Manuals 5.5 Westinghouse STS5.5-16Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) [OPTION B] a.A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, asmodified by the following exceptions:1.The visual examination of containment concrete surfaces intended tofulfill the requirements of 10 CFR 50, Appendix J, Option B testing,will be performed in accordance with the requirements of andfrequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC. 2.The visual examination of the steel liner plate inside containmentintended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC. [ 3. . . . ] b.The calculated peak containment internal pressure for the design basis lossof coolant accident, Pa, is [45 psig]. The containment design pressure is[50 psig].c.The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% ofcontainment air weight per day. d.Leakage rate acceptance criteria are:1.Containment leakage rate acceptance criterion is 1.0 La. During thefirst unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and 0.75 La for Type A tests. 2.Air lock testing acceptance criteria are:a)Overall air lock leakage rate is [0.05 La] when tested at Pa.b)For each door, leakage rate is [0.01 La] when pressurized to[ 10 psig].14 0.2511.33INSERT 86.8.4.h 6.8.4.h 6.8.4.h 6.8.4.h 6.8.4.h.a 6.8.4.h.b 6.8.4.h.b.1 6.8.4.h.b.2 6.8.4.h 6 INSERT 9 for at least two minutes103422233 3312 5.5 Insert Page 5.5-16 CTS INSERT 8 1.Bypass leakage paths to the auxiliary building leakage from isolation valves that are sealedwith fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal system capacity isadequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.INSERT 9 3.For each containment purge supply and exhaust isolation valve, acceptance criteria ismeasured leakage rate to 0.05 La.4.Bypass leakage paths to the auxiliary building acceptance criteria are:a)The combined bypass leakage rate to the auxiliary building shall be 0.25 La byapplicable Type B and C tests.b)Penetrations not individually testable shall have no detectable leakage when tested withsoap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type Atest.6.8.4.h 6.8.4.h.d 6.8.4.h.d.1 6.8.4.h.d.2 6.8.4.h.c 33 Programs and Manuals 5.5 Westinghouse STS5.5-17Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) e.The provisions of SR 3.0.3 are applicable to the Containment Leakage RateTesting Program.f.Nothing in these Technical Specifications shall be construed to modify thetesting Frequencies required by 10 CFR 50, Appendix J. [OPTION A/B Combined] a.A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A] test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions: 1.The visual examination of containment concrete surfaces intended tofulfill the requirements of 10 CFR 50, Appendix J, Option B testing,will be performed in accordance with the requirements of andfrequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC. 2.The visual examination of the steel liner plate inside containmentintended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC. [ 3. . . . ] b.The calculated peak containment internal pressure for the design basis lossof coolant accident, Pa, [45 psig]. The containment design pressure is [50 psig]. c.The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% ofcontainment air weight per day. d.Leakage rate acceptance criteria are:14 6.8.4.h 6.8.4.h 6.8.4.h 104 Programs and Manuals 5.5 Westinghouse STS5.5-18Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) 1.Containment leakage rate acceptance criterion is 1.0 La. During thefirst unit startup following testing in accordance with this program, theleakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests]. 2.Air lock testing acceptance criteria are:a)Overall air lock leakage rate is [0.05 La] when tested at Pa.b)For each door, leakage rate is [0.01 La] when pressurized to[ 10 psig]. e.The provisions of SR 3.0.3 are applicable to the Containment Leakage RateTesting Program. f.Nothing in these Technical Specifications shall be construed to modify thetesting Frequencies required by 10 CFR 50, Appendix J. 5.5.17 Battery Monitoring and Maintenance Program --------------------------------------REVIEWER'S NOTE------------------------------------------ This program and the corresponding requirements in LCO 3.8.4, LCO 3.8.5, and LCO 3.8.6 require providing the information and verifications requested in the Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite - Update to TSTF-360," (76FR54510). ------------------------------------------------------------------------------------------------------------ This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below: a.The program allows the following RG 1.129, Revision 2 exceptions:1.Battery temperature correction may be performed before or afterconducting discharge tests.2.RG 1.129, Regulatory Position 1, Subsection 2, "References," is notapplicable to this program.DOC M03 14 15 10445 Programs and Manuals 5.5 Westinghouse STS5.5-19Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.17 Battery Monitoring and Maintenance Program (continued) 3.In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2,"Inspections," the following shall be used: "Where reference is madeto the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery."4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5.In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6,"Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string."b.The program shall include the following provisions:1.Actions to restore battery cells with float voltage < [2.13] V;2.Actions to determine whether the float voltage of the remainingbattery cells is [2.13] V when the float voltage of a battery cell hasbeen found to be < [2.13] V;3.Actions to equalize and test battery cells that had been discoveredwith electrolyte level below the top of the plates;4.Limits on average electrolyte temperature, battery connectionresistance, and battery terminal voltage; and5.A requirement to obtain specific gravity readings of all cells at eachdischarge test, consistent with manufacturer recommendations.5.5.18 Control Room Envelope (CRE) Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of [5 rem whole body or its equivalent to any part of the body] [5 rem total 22DOC M01 15 16 6.17 244Table 4.8.2 Float Voltage Ventilation V3 Programs and Manuals 5.5 Westinghouse STS5.5-20Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.18 Control Room Envelope (CRE) Habitability Program (continued) effective dose equivalent (TEDE)] for the duration of the accident. The program shall include the following elements: a.The definition of the CRE and the CRE boundary. b.Requirements for maintaining the CRE boundary in its design conditionincluding configuration control and preventive maintenance.c.Requirements for (i) determining the unfiltered air inleakage past the CREboundary into the CRE in accordance with the testing methods and at theFrequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear PowerReactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.[The following are exceptions to Sections C.1 and C.2 of RegulatoryGuide 1.197, Revision 0: 1.;and]d.Measurement, at designated locations, of the CRE pressure relative to allexternal areas adjacent to the CRE boundary during the pressurizationmode of operation by one train of the CREFS, operating at the flow rate required by the VFTP, at a Frequency of [18] months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the [18] month assessment of the CRE boundary. e.The quantitative limits on unfiltered air inleakage into the CRE. These limitsshall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfilteredair inleakage limits for hazardous chemicals must ensure that exposure ofCRE occupants to these hazards will be within the assumptions in the licensing basis.f.The provisions of SR 3.0.2 are applicable to the Frequencies for assessingCRE habitability, determining CRE unfiltered inleakage, and measuringCRE pressure and assessing the CRE boundary as required by paragraphsc and d, respectively.16 26.17.a 6.17.b 6.17.c 6.17.d 26.17.e 6.17.f42226.17 V3 Programs and Manuals 5.5 Westinghouse STS5.5-21Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.19 [ Setpoint Control Program This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analyses, provides a means for processing changes to instrumentation setpoints, and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verifies that instrumentation will function as required. a.The program shall list the Functions in the following specifications to which itapplies: 1.LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation;"2.LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)Instrumentation Functions;"3.LCO 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) StartInstrumentation;"4.LCO 3.3.6, "Containment Purge and Exhaust IsolationInstrumentation;"5.LCO 3.3.7, "Control Room Emergency Filtration System (CREFS)Actuation Instrumentation;"6.LCO 3.3.8, "Fuel Building Air Cleanup System (FBACS) ActuationInstrumentation;" and7.LCO 3.3.9, "Boron Dilution Protection System (BDPS)."b.The program shall require the Nominal Trip Setpoint (NTSP), AllowableValue (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in paragraph a. are calculated using the NRC approved setpoint methodology, as listed below. In addition, the program shall contain the value of the NTSP, AV, AFT, and ALT (as applicable) for each Function described in paragraph a. and shall identify the setpoint methodology used to calculate these values. ---------------------------------------Reviewer's Note---------------------------------------- List the NRC safety evaluation report by letter, date, and ADAMS accession number (if available) that approved the setpoint methodologies. ---------------------------------------------------------------------------------------------------- 1.[Insert reference to NRC safety evaluation that approved the setpointmethodology.] c.The program shall establish methods to ensure that Functions described inparagraph a. will function as required by verifying the as-left and as-foundsettings are consistent with those established by the setpoint methodology.4 Programs and Manuals 5.5 Westinghouse STS5.5-22Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued) d.-----------------------------------REVIEWER'S NOTE--------------------------------------A license amendment request to implement a Setpoint Control Programmust list the instrument functions to which the program requirements ofparagraph d. will be applied. Paragraph d. shall apply to all Functions in the Reactor Trip System and Engineered Safety Feature Actuation System specifications unless one or more of the following exclusions apply: 1.Manual actuation circuits, automatic actuation logic circuits or toinstrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded. 2.Settings associated with safety relief valves are excluded. Theperformance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program. 3.Functions and Surveillance Requirements which test only digitalcomponents are normally excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply. ------------------------------------------------------------------------------------------------------- The program shall identify the Functions described in paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The NTSP of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, CHANNEL OPERATIONAL TESTS, and TRIP ACTUATING DEVICE OPERATIONAL TESTS that verify the NTSP. 1 The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified NTSP. 2.If the as-found value of the instrument channel trip setting differs fromthe previous as-left value or the specified NTSP by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then theinstrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program. 4 Programs and Manuals 5.5 Westinghouse STS 5.5-23 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX35.5 Programs and Manuals

5.5.19 Setpoint Control Program (continued)

3. If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

4. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NTSP at the completion of the surveillance test; otherwise, the channel is inoperable (setpoints may be more conservative than the NTSP provided that the as-found and as-left tolerances apply to the actual setpoint used to confirm channel performance). e. The program shall be specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference]. ] 5.5.20 [ Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. ] 17 DOC M04 4422 Programs and Manuals 5.5 Westinghouse STS5.5-1Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM) a.The ODCM shall contain the methodology and parameters used in thecalculation of offsite doses resulting from radioactive gaseous and liquideffluents, in the calculation of gaseous and liquid effluent monitoring alarmand trip setpoints, and in the conduct of the radiological environmental monitoring program, andb.The ODCM shall also contain the radioactive effluent controls andradiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.1] and Specification [5.6.2]. Licensee initiated changes to the ODCM: a.Shall be documented and records of reviews performed shall be retained.This documentation shall contain:1.Sufficient information to support the change(s) together with theappropriate analyses or evaluations justifying the change(s) and2.A determination that the change(s) maintain the levels of radioactiveeffluent control required by 10 CFR 20.1302, 40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations,b.Shall become effective after the approval of the plant manager, andc.Shall be submitted to the NRC in the form of a complete, legible copy of theentire ODCM as a part of or concurrent with the Radioactive EffluentRelease Report for the period of the report in which any change in theODCM was made. Each change shall be identified by markings in themargin of the affected pages, clearly indicating the area of the page thatwas changed, and shall indicate the date (i.e., month and year) the change was implemented.1.18 1.18 1.18 6.14.1.1 6.14.1.1.a 6.14.1.1.b 6.14.1.1 6.14.1.2 6.14.1.3 2; 1 Programs and Manuals 5.5 Westinghouse STS 5.5-2 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include [Recirculation Spray, Safety Injection, Chemical and Volume Control, gas stripper, and Hydrogen Recombiner]. The program shall include the following: a. Preventive maintenance and periodic visual inspection requirements and

b. Integrated leak test requirements for each system at least once per [18] months. The provisions of SR 3.0.2 are applicable.

5.5.3 [ Post Accident Sampling

----------------------------------------REVIEWER'S NOTE---------------------------------------- This program may be eliminated based on the implementation of WCAP-14986, Rev. 1, "Post Accident Sampling System Requirements: A Technical Basis," and the associated NRC Safety Evaluation dated June 14, 2000, and implementation of the following commitments: 1. [Licensee] has developed contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere. The contingency plans will be contained in emergency plan implementing procedures and implemented with the implementation of the License amendment. Establishment of contingency plans is considered a regulatory commitment. 2. The capability for classifying fuel damage events at the Alert level threshold has been established for [Plant] at radioactivity levels of 300 mCi/cc dose equivalent iodine. This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations. This capability will be described in emergency plan implementing procedures and implemented with the implementation of the License amendment. The capability for classifying fuel damage events is considered a regulatory commitment. 3. [Licensee] has established the capability to monitor radioactive iodines that have been released to offsite environs. This capability is described in our emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment. ------------------------------------------------------------------------------------------------------------

This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents 6.8.4.a 6.8.4.a 6.8.4.b 2Residual Heat Removal, Containment Spray, and RCS Sampling 4 Programs and Manuals 5.5 Westinghouse STS 5.5-3 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.3 Post Accident Sampling (continued) and containment atmosphere samples under accident conditions. The program shall include the following: a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analysis equipment. ]

5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I,

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, 3 6.8.4.f 6.8.4.f 6.8.4.f.1) 6.8.4.f.2) 6.8.4.f.3) 6.8.4.f.4) 6.8.4.f.5) 6.8.4.f.6) 44 Programs and Manuals 5.5 Westinghouse STS 5.5-4 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.4 Radioactive Effluent Controls Program (continued)

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ, h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I,

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the FSAR, Section [ ], cyclic and transient occurrences to ensure that components are maintained within the design limits. 5.5.6 [ Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC. The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ] 3 4 6.8.4.f.7) 6.8.4.f.7)).1 6.8.4.f.7)).2 6.8.4.f.8) 6.8.4.f.9) 6.8.4.f.10) 6.8.4.f 6.8.4.f 6.8.4.l 5.2.1 4244U Programs and Manuals 5.5 Westinghouse STS5.5-5Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals. ---------------------------------------REVIEWER'S NOTE---------------------------------------- The inspection interval and scope for RCP flywheels stated above can be applied to plants that satisfy the requirements in WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination." ------------------------------------------------------------------------------------------------------------ 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following: a.Testing frequencies applicable to the ASME Code for Operations andMaintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Required Frequencies for performing inservice testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days b.The provisions of SR 3.0.2 are applicable to the above requiredFrequencies and to other normal and accelerated Frequencies specified as2 years or less in the Inservice Testing Program for performing inservicetesting activities,c.The provisions of SR 3.0.3 are applicable to inservice testing activities, and6 5 4.0.5.c 4.0.5 4.0.5 4.0.5.b 4.0.5.c 4.0.5.d 544pumps and valves 6ultrasonic liquid penetrantmagnetic particle 33 Programs and Manuals 5.5 Westinghouse STS5.5-6Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.8 Inservice Testing Program (continued) d.Nothing in the ASME OM Code shall be construed to supersede therequirements of any TS.5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: a.Provisions for condition monitoring assessments. Condition monitoringassessment means an evaluation of the "as found" condition of the tubingwith respect to the performance criteria for structural integrity and accidentinduced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging [or repair] oftubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, [or repaired] toconfirm that the performance criteria are being met.b.Performance criteria for SG tube integrity. SG tube integrity shall bemaintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1.Structural integrity performance criterion: All in-service steamgenerator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transientsincluded in the design specification) and design basis accidents. Thisincludes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated withthe design basis accidents, or combination of accidents in accordancewith the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2on the combined primary loads and 1.0 on axial secondary loads.2.Accident induced leakage performance criterion: The primary tosecondary accident induced leakage rate for any design basisaccident, other than a SG tube rupture, shall not exceed the leakagerate assumed in the accident analysis in terms of total leakage rate forall SGs and leakage rate for an individual SG. Leakage is not to), , 6.8.4.k 6.8.4.k.a 6.8.4.k.b 6.8.4.k.b.1 6.8.4.k.b.2 22TSTF-510-A6 7 4.0.5.e 4.0.5 44or 1 Programs and Manuals 5.5 Westinghouse STS 5.5-7 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued) exceed [1 gpm] per SG [, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding [40%] of the nominal tube wall thickness shall be plugged [or repaired]. ---------------------------------------REVIEWER'S NOTE---------------------------------------- Alternate tube repair criteria currently permitted by plant technical specifications are listed here. The description of these alternate tube repair criteria should be equivalent to the descriptions in current technical specifications and should also include any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria. --------------------------------------------------------------------------------------------------------------- [The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. . . .] d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. 6.8.4.k.b.3 6.8.4.k.c 6.8.4.k.d ] plugging [or ]] plugging [or ]]plugging [or plugging [or ]plugging [or ] assessment 22TSTF-510-ATSTF-510-A5TSTF-510-ATSTF-510-A27 4TSTF-510-A7 Programs and Manuals 5.5 Westinghouse STS5.5-8Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) ---------------------------------------REVIEWER'S NOTE---------------------------------------- Plants are to include the appropriate Frequency (e.g., select the appropriate Item 2.) for their SG design. The first Item 2 is applicable to SGs with Alloy 600 mill annealed tubing. The second Item 2 is applicable to SGs with Alloy 600 thermally treated tubing. The third Item 2 is applicable to SGs with Alloy 690 thermally treated tubing. ------------------------------------------------------------------------------------------------------------ 1.Inspect 100% of the tubes in each SG during the first refueling outagefollowing SG replacement. [2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.] [2. Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.] [2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.] 3.If crack indications are found in any SG tube, then the next inspectionfor each SG for the degradation mechanism that caused the crackindication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such asfrom examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.e.Provisions for monitoring operational primary to secondary LEAKAGE.INSERT 3 INSERT 1 INSERT 2 installationaffected and potentially affected results in more frequent inspections 6.8.4.k.d.1 6.8.4.k.d.2 6.8.4.k.d.3 6.8.4.k.e 2TSTF-510-ATSTF-510-A7 54 5.5 Insert Page 5.5-8 CTS INSERT 1 After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. INSERT 2 After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. ------------------------------------------------- Reviewer's Note ------------------------------------------------------ A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods. ------------------------------------------------------------------------------------------------------------------------------- a)After the first refueling outage following SG installation, inspect 100% of the tubes during thenext 120 effective full power months. This constitutes the first inspection period; 2TSTF-510-A2 5.5 Insert Page 5.5-8 CTS b)During the next 96 effective full power months, inspect 100% of the tubes. This constitutesthe second inspection period; and c)During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full powermonths. This constitutes the third and subsequent inspection periods. INSERT 3 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. ----------------------------------------------- Reviewer's Note ------------------------------------------------------ A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods. ----------------------------------------------------------------------------------------------------------------------------- a)After the first refueling outage following SG installation, inspect 100% of the tubes during thenext 144 effective full power months. This constitutes the first inspection period;b)During the next 120 effective full power months, inspect 100% of the tubes. This constitutesthe second inspection period;c)During the next 96 effective full power months, inspect 100% of the tubes. This constitutesthe third inspection period; andd)During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full powermonths. This constitutes the fourth and subsequent inspection periods.2TSTF-510-A Programs and Manuals 5.5 Westinghouse STS 5.5-9 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

  [f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.  ---------------------------------------REVIEWER'S NOTE---------------------------------------- Tube repair methods currently permitted by plant technical specifications are to be listed here. The description of these tube repair methods should be equivalent to the descriptions in current technical specifications. If there are no approved tube repair methods, this section should not be used. ------------------------------------------------------------------------------------------------------------

1. . . .]

5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables,

b. Identification of the procedures used to measure the values of the critical variables,

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage,

d. Procedures for the recording and management of data,

e. Procedures defining corrective actions for all off control point chemistry conditions, and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. 27 8 6.8.4.c 6.8.4.c 6.8.4.c.(i) 6.8.4.c.(ii) 6.8.4.c.(iii) 6.8.4.c.(iv) 6.8.4.c.(vi) 6.8.4.c.(v) 48452; ; ; ; ; 11 Programs and Manuals 5.5 Westinghouse STS5.5-10Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, Revision 2, ASME N510-1989, and AG-1]. a.Demonstrate for each of the ESF systems that an inplace test of the highefficiency particulate air (HEPA) filters shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%]. ESF Ventilation System Flowrate [ ] [ ] b.Demonstrate for each of the ESF systems that an inplace test of thecharcoal adsorber shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%]. ESF Ventilation System Flowrate [ ] [ ] c.Demonstrate for each of the ESF systems that a laboratory test of a sampleof the charcoal adsorber, when obtained as described in [Regulatory Guide 1.52, Revision 2], shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity specified below. ESF Ventilation System Penetration RH Face Velocity (fps) [ ] [See Reviewer's [See [See Reviewer's Note] Reviewer's Note] Note] ----------------------------------------REVIEWER'S NOTE---------------------------------------- The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate. 9 INSERT 5 ANSI N510-1975 (except for the provisions of Sections 8 and 9)ANSI N510-1975Regulatory Positions C.5.a, C.5.c, C.5.d and C.6.b of ASTM D3803-19894.6.1.8.e 4.7.7.c.3 4.7.7.f 4.7.8.b.3 4.7.8.e 4.9.12.b.3 4.9.12.e removal efficiently of 99.95% of a halogenated hydrocarbon refrigerant test gas removal efficiently of 99.95% of dioctyl phthalate (DOP) ANSI N510-1975 (except for the provisions of Sections 8 and 9)4.6.1.8.f 4.7.7.c.3 4.7.7.g 4.7.8.b.3 4.7.8.f 4.9.12.b.3 4.9.12.f INSERT 6 Regulatory Position C.6.b of of 70% < 2.5%4.6.1.8.b.2 4.6.1.8.c 4.7.7.c..2 4.7.7.d 4.7.8.b.2 4.7.8.c 4.9.12.b.2 4.9.12.c EGTS ABGTS CREVS Regulatory Positions C.5.a and C.5.c of Regulatory Positions C.5.a and C.5.d of4.6.1.8.b.1 4.7.7.c.1 4.7.8.b.1 4.9.12.b.1 4323322232323323325INSERT 4 3 5.5 Insert Page 5.5-10 CTS INSERT 4 The test described in Specification 5.5.9.a and 5.5.9.b shall be performed once per 18 months; after any structural maintenance on the high efficiency particulate air (HEPA) filter bank or charcoal adsorber bank housing; following painting, fire, or chemical release in any ventilation zone communicating with the system; and after each complete or partial replacement of a HEPA filter bank or charcoal adsorber bank. The test described in Specification 5.5.9.c shall be performed once per 18 months or after 720 hours of filter operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and following painting, fire, or chemical release in any ventilation zone communicating with the system. The test described in Specification 5.5.9.d and 5.5.9.e shall be performed once per 18 months. INSERT 5 ESF Ventilation System Flow Rate (cfm) Emergency Gas Treatment System (EGTS) 4000 Auxiliary Building Gas Treatment System (ABGTS) 9000 Control Room Emergency Ventilation System (CREVS) 4000 INSERT 6 ESF Ventilation System Flow Rate (cfm) EGTS 4000 ABGTS 9000 CREVS 4000 3334.6.1.8.b 4.6.1.8.e 4.6.1.8.f 4.7.7.f 4.7.7.g 4.7.8.e 4.7.8.f 4.6.1.8.c 4.6.1.8.b 4.7.7.c 4.7.7.d 4.7.8.b 4.7.8.c 4.9.12.b 4.9.12.c 4.6.1.8.d 4.7.7.e 4.7.8.d 4.9.12.d Programs and Manuals 5.5 Westinghouse STS5.5-11Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued) ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95% (or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test. Allowable Penetration = [(100% - Methyl Iodide Efficiency

for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor] When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following: Safety factor 2 for systems with or without humidity control. Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions. If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified. *This value should be the efficiency that was incorporated in the licensee'saccident analysis which was reviewed and approved by the staff in a safety evaluation. ------------------------------------------------------------------------------------------------------------ d.Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters, the prefilters, and the charcoal adsorbers is lessthan the value specified below when tested in accordance with [RegulatoryGuide 1.52, Revision 2, and ASME N510-1989] at the system flowratespecified below [+/- 10%].ESF Ventilation System Delta P Flowrate [ ] [ ] [ ] [ e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below [+/- 10%] when tested in accordance with [ASME N510-1989]. ESF Ventilation System Wattage ] [ ] [ ] The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 9 INSERT 7 4.6.1.8.d.1 4.7.7.e.1 4.7.8.b.3 4.7.8.d.1 4.9.12.b.3 4.9.12.d.1 4.7.7.c.3 ANSI N510-1975 Auxiliary Building Gas Treatment System32 kW4.7.8.d.4 4.9.12.d.3 5323241975 5.5 Insert Page 5.5-11 CTS INSERT 7 ESF Ventilation System Combined Delta P (inches water gauge) Flowrate (cfm) EGTS54000ABGTS39000CREVS340003 Programs and Manuals 5.5 Westinghouse STS5.5-12Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous radioactivity quantities shall be determined following the methodology in [Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be determined in accordance with [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"]. The program shall include: a.The limits for concentrations of hydrogen and oxygen in the [Waste GasHoldup System] and a surveillance program to ensure the limits aremaintained. Such limits shall be appropriate to the system's design criteria(i.e., whether or not the system is designed to withstand a hydrogen explosion),b.A surveillance program to ensure that the quantity of radioactivity containedin [each gas storage tank and fed into the offgas treatment system] is lessthan the amount that would result in a whole body exposure of 0.5 rem toany individual in an unrestricted area, in the event of [an uncontrolledrelease of the tanks' contents], andc.A surveillance program to ensure that the quantity of radioactivity containedin all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the [Liquid Radwaste Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. 10 3.11.1.4 3.11.2.5 3.11.2.6 the Condensate Storage Tank, Steam Generator Layup Tank, and 4.11.2.5 4.11.2.6 4.11.1.4 storage temporary temporary29422decay decay , Condensate Storage Tank, and Steam Generator Layup Tank 32exceeding ; 11; 11 Programs and Manuals 5.5 Westinghouse STS5.5-13Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following: a.Acceptability of new fuel oil for use prior to addition to storage tanks bydetermining that the fuel oil has:1.An API gravity or an absolute specific gravity within limits,2.A flash point and kinematic viscosity within limits for ASTM 2D fuel oil,and3.A clear and bright appearance with proper color or a water andsediment content within limits.b.Within 31 days following addition of the new fuel oil to storage tanks, verifythat the properties of the new fuel oil, other than those addressed in a.,above, are within limits for ASTM 2D fuel oil, andc.Total particulate concentration of the fuel oil is 10 mg/l when tested every31 days.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies. 5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications. a.Changes to the Bases of the TS shall be made under appropriateadministrative controls and reviews.b.Licensees may make changes to Bases without prior NRC approvalprovided the changes do not require either of the following:1.A change in the TS incorporated in the license or2.A change to the updated FSAR or Bases that requires NRC approvalpursuant to 10 CFR 50.59.6.16 1112 6.16 6.16.a 6.16.a.1 6.16.a.2 6.16.a.3 6.16.b 6.16.c 6.8.4.j 6.8.4.j 6.8.4.j.a 6.8.4.j.b 6.8.4.j.b.1 6.8.4.j.b.2 44 Programs and Manuals 5.5 Westinghouse STS5.5-14Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.14 Technical Specifications (TS) Bases Control Program (continued) c.The Bases Control Program shall contain provisions to ensure that theBases are maintained consistent with the FSAR.d.Proposed changes that meet the criteria of Specification 5.5.14b above shallbe reviewed and approved by the NRC prior to implementation. Changes tothe Bases implemented without prior NRC approval shall be provided to theNRC on a frequency consistent with 10 CFR 50.71(e).5.5.15 Safety Function Determination Program (SFDP) This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following: a.Provisions for cross train checks to ensure a loss of the capability to performthe safety function assumed in the accident analysis does not go undetected,b.Provisions for ensuring the plant is maintained in a safe condition if a loss offunction condition exists, c.Provisions to ensure that an inoperable supported system's CompletionTime is not inappropriately extended as a result of multiple support system inoperabilities, and d.Other appropriate limitations and remedial or compensatory actions.A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: a.A required system redundant to the system(s) supported by the inoperablesupport system is also inoperable, or b.A required system redundant to the system(s) in turn supported by theinoperable supported system is also inoperable, or12 13 6.8.4.j.d 6.8.4.j.d 6.8.4.j.d DOC M02 4412 4U3; 11; ; ; ; ;

Programs and Manuals 5.5 Westinghouse STS 5.5-15 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.15 Safety Function Determination Program (SFDP) (continued)

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system. 5.5. 16 Containment Leakage Rate Testing Program

  [OPTION A]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

b. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

c. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests. 2. Air lock testing acceptance criteria are: a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

b) For each door, leakage rate is [0.01 La] when pressurized to [ 10 psig].

d. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

e. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

13 14 DOC M02 4410 Programs and Manuals 5.5 Westinghouse STS5.5-16Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) [OPTION B] a.A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, asmodified by the following exceptions:1.The visual examination of containment concrete surfaces intended tofulfill the requirements of 10 CFR 50, Appendix J, Option B testing,will be performed in accordance with the requirements of andfrequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC. 2.The visual examination of the steel liner plate inside containmentintended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC. [ 3. . . . ] b.The calculated peak containment internal pressure for the design basis lossof coolant accident, Pa, is [45 psig]. The containment design pressure is[50 psig].c.The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% ofcontainment air weight per day. d.Leakage rate acceptance criteria are:1.Containment leakage rate acceptance criterion is 1.0 La. During thefirst unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and 0.75 La for Type A tests. 2.Air lock testing acceptance criteria are:a)Overall air lock leakage rate is [0.05 La] when tested at Pa.b)For each door, leakage rate is [0.01 La] when pressurized to[ 10 psig].14 0.2511.33INSERT 86.8.4.h 6.8.4.h 6.8.4.h 6.8.4.h 6.8.4.h.a 6.8.4.h.b 6.8.4.h.b.1 6.8.4.h.b.2 6.8.4.h 6 INSERT 9 for at least two minutes103422233 3312 5.5 Insert Page 5.5-16 CTS INSERT 8 1.Bypass leakage paths to the auxiliary building leakage from isolation valves that are sealedwith fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal system capacity isadequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.INSERT 9 3.For each containment purge supply and exhaust isolation valve, acceptance criteria ismeasured leakage rate to 0.05 La.4.Bypass leakage paths to the auxiliary building acceptance criteria are:a)The combined bypass leakage rate to the auxiliary building shall be 0.25 La byapplicable Type B and C tests.b)Penetrations not individually testable shall have no detectable leakage when tested withsoap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type Atest.6.8.4.h 6.8.4.h.d 6.8.4.h.d.1 6.8.4.h.d.2 6.8.4.h.c 33 Programs and Manuals 5.5 Westinghouse STS 5.5-17 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.16 Containment Leakage Rate Testing Program (continued)

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J. [OPTION A/B Combined] a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A] test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

   [ 3. . . . ]      b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, [45 psig]. The containment design pressure is [50 psig].

c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

d. Leakage rate acceptance criteria are:

14 6.8.4.h 6.8.4.h 6.8.4.h 104 Programs and Manuals 5.5 Westinghouse STS 5.5-18 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.16 Containment Leakage Rate Testing Program (continued)

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests]. 2. Air lock testing acceptance criteria are: a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

b) For each door, leakage rate is [0.01 La] when pressurized to [ 10 psig].

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.17 Battery Monitoring and Maintenance Program

  --------------------------------------REVIEWER'S NOTE------------------------------------------    This program and the corresponding requirements in LCO 3.8.4, LCO 3.8.5, and LCO 3.8.6 require providing the information and verifications requested in the Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite - Update to TSTF-360," (76FR54510).    ------------------------------------------------------------------------------------------------------------  This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below:     a. The program allows the following RG 1.129, Revision 2 exceptions:

1. Battery temperature correction may be performed before or after conducting discharge tests.

2. RG 1.129, Regulatory Position 1, Subsection 2, "References," is not applicable to this program. DOC M03 14 15 10445 Programs and Manuals 5.5 Westinghouse STS 5.5-19 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.17 Battery Monitoring and Maintenance Program (continued)

3. In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2, "Inspections," the following shall be used: "Where reference is made to the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery."

4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage < [2.13] V;

2. Actions to determine whether the float voltage of the remaining battery cells is [2.13] V when the float voltage of a battery cell has been found to be < [2.13] V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates; 4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

5.5.18 Control Room Envelope (CRE) Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of [5 rem whole body or its equivalent to any part of the body] [5 rem total 22DOC M01 15 16 6.17 244Table 4.8.2 Float Voltage Ventilation V3 Programs and Manuals 5.5 Westinghouse STS 5.5-20 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.18 Control Room Envelope (CRE) Habitability Program (continued) effective dose equivalent (TEDE)] for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

   [The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. ;and]

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFS, operating at the flow rate required by the VFTP, at a Frequency of [18] months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the

[18] month assessment of the CRE boundary. e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively. 16 26.17.a 6.17.b 6.17.c 6.17.d 26.17.e 6.17.f 42226.17 V3 Programs and Manuals 5.5 Westinghouse STS 5.5-21 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.19 [ Setpoint Control Program This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analyses, provides a means for processing changes to instrumentation setpoints, and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verifies that instrumentation will function as required.

a. The program shall list the Functions in the following specifications to which it applies:

1. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation;" 2. LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation Functions;" 3. LCO 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation;" 4. LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation;" 5. LCO 3.3.7, "Control Room Emergency Filtration System (CREFS) Actuation Instrumentation;" 6. LCO 3.3.8, "Fuel Building Air Cleanup System (FBACS) Actuation Instrumentation;" and 7. LCO 3.3.9, "Boron Dilution Protection System (BDPS)."

b. The program shall require the Nominal Trip Setpoint (NTSP), Allowable Value (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in paragraph a. are calculated using the NRC approved setpoint methodology, as listed below. In addition, the program shall contain the value of the NTSP, AV, AFT, and ALT (as applicable) for each Function described in paragraph a. and shall identify the setpoint methodology used to calculate these values. ---------------------------------------Reviewer's Note---------------------------------------- List the NRC safety evaluation report by letter, date, and ADAMS accession number (if available) that approved the setpoint methodologies. - --------------------------------------------------------------------------------------------------- 1. [Insert reference to NRC safety evaluation that approved the setpoint methodology.] c. The program shall establish methods to ensure that Functions described in paragraph a. will function as required by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology. 4 Programs and Manuals 5.5 Westinghouse STS 5.5-22 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals

5.5.19 Setpoint Control Program (continued)

d. -----------------------------------REVIEWER'S NOTE-------------------------------------- A license amendment request to implement a Setpoint Control Program must list the instrument functions to which the program requirements of paragraph d. will be applied. Paragraph d. shall apply to all Functions in the Reactor Trip System and Engineered Safety Feature Actuation System specifications unless one or more of the following exclusions apply:

1. Manual actuation circuits, automatic actuation logic circuits or to instrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded.

2. Settings associated with safety relief valves are excluded. The performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program.

3. Functions and Surveillance Requirements which test only digital components are normally excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply. ------------------------------------------------------------------------------------------------------- The program shall identify the Functions described in paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The NTSP of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, CHANNEL OPERATIONAL TESTS, and TRIP ACTUATING DEVICE OPERATIONAL TESTS that verify the NTSP.

1 The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified NTSP.

2. If the as-found value of the instrument channel trip setting differs from the previous as-left value or the specified NTSP by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then the instrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program. 4 Programs and Manuals 5.5 Westinghouse STS5.5-23Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX35.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued) 3.If the as-found value of the instrument channel trip setting is lessconservative than the specified AV, then the SR is not met and theinstrument channel shall be immediately declared inoperable. 4.The instrument channel setpoint shall be reset to a value that is withinthe as-left tolerance around the NTSP at the completion of thesurveillance test; otherwise, the channel is inoperable (setpoints maybe more conservative than the NTSP provided that the as-found andas-left tolerances apply to the actual setpoint used to confirm channelperformance).e.The program shall be specified in [insert the facility FSAR reference or thename of any document incorporated into the facility FSAR by reference]. ] 5.5.20 [ Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. a.The Surveillance Frequency Control Program shall contain a list ofFrequencies of those Surveillance Requirements for which the Frequency is controlled by the program.b.Changes to the Frequencies listed in the Surveillance Frequency ControlProgram shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.c.The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicableto the Frequencies established in the Surveillance Frequency Control Program. ]17 DOC M04 4422 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and Unit 2 Page 1 of 1 1.Typographical/grammatical error corrected.2.The ISTS contains bracketed information and/or values that are generic toWestinghouse vintage plants. The brackets are removed and the proper plantspecific information/value is inserted to reflect the current licensing basis.3.Changes are made (additions, deletions, and/or changes) to the ISTS that reflectthe plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.4.The bracketed ISTS 5.5.3, "Post Accident Sampling," ISTS 5.5.6, "Pre-StressedConcrete Containment Tendon Surveillance Program," and ISTS 5.5.19, "Setpoint Control Program," are not included in the SQN ITS. Subsequent programs in the ITS Section 5.5 have been renumbered, as necessary.5.The Reviewer's Note has been deleted. This information is for the NRC reviewerto be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.6.The Inservice Testing (IST) Program (ISTS 5.5.8) has been modified to state thatthe IST Program provides control for ASME Code Class 1, 2, and 3 "pumps and valves" in place of the current "components." 10 CFR 50.55a(f) provides theregulatory requirements for an IST Program. It specifies that ASME Code Class 1,2, and 3 pumps and valves are the only components covered by an IST Program.

10 CFR 50.55a(g) provides regulatory requirements for an Inservice Inspection (ISI) Program. It specifies that ASME Code Class 1, 2, and 3 components are covered by the ISI Program, and that pumps and valves are covered by the ISTProgram in 10 CFR 50.55a(f). The ISTS does not include ISI Program requirements as these requirements have been relocated to a plant specific document. Therefore, the components to which the IST Program applies (i.e., pumps and valves) have been added for clarity.7.Changes made to improve clarity.8.ISTS 5.5.10 (ITS 5.5.8) provides the requirements for the Secondary WaterChemistry Program. The program in the ISTS includes requirements to providecontrols for monitoring secondary water chemistry to inhibit SG tube degradationand low pressure turbine disc stress corrosion. ITS 5.5.8 provides controls formonitoring secondary water chemistry only to inhibit SG tube degradation. Thischange is consistent with the current SQN licensing bases.9.The program details of the Explosive Gas and Storage Tank RadioactivityMonitoring Program are described in ISTS 5.5.12 (ITS 5.5.10) parts a, b, and c. Therefore, the sentence in the introductory paragraph that specifies a method todetermine the explosive gas and storage tank radioactivity is not necessary.10.SQN complies with Option B of 10 CFR 50, Appendix J. Therefore, the ISTS 5.5.16Option A and combined Option A and B provisions have been deleted.11.These punctuation corrections have been made consistent with the Writers Guidefor the Improved Standard Technical Specifications NEI 01-03, Section 5.1.3.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 6 ITS 5.6, REPORTING REQUIREMENTS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 5.6 6.0 ADMINISTRATIVE CONTROLS 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary-to-secondary leakage. l. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits. 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4. STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS 1/ 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. 6.9.1.5 DELETED

_________________ 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. August 2, 2006 SEQUOYAH - UNIT 1 6-11b Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306, 309 5.6 5.6 A02A02Page 1 of 16 See ITS 5.5 A01ITS ITS 5.6 ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 1/ 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. 6.9.1.7 (Relocated to the ODCM.) ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1/ 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

6.9.1.9 (Relocated to the ODCM or PCP.) ____________________ 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

April 5, 2005 SEQUOYAH - UNIT 1 6-12 Amendment No. 42, 58, 74, 117, 148, 169, 281, 300 by May 15L01INSERT 1M015.6.1 5.6.2 5.6.1 Note, 5.6.2 Note Page 2 of 16 in accordance with 10 CFR 50.36aA03 ITS 5.6 Insert Page 6-12 INSERT 1 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

M01Page 3 of 16 A01ITS ITS 5.6 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT6.9.1.10 DELETED. CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or anyremaining part of a reload cycle for the following: 1.f1(I) limits for Overtemperature Delta T Trip Setpoints and f2(I) limits for Overpower DeltaT Trip Setpoints for Specification 2.2.1.2.Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit forSpecification 3/4.1.1.3,3.Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,4.Control Bank Insertion Limits for Specification 3/4.1.3.6,5.AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,6.Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and7.Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents: The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). 1.BAW-10180P-A, "NEMO - Nodal Expansion Method Optimized"2.BAW-10169P-A, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology forRecirculating Steam Generator Plants" 3.BAW-10163P-A, "Core Operating Limit Methodology for Westinghouse-Designed PWRs"4.BAW-10168P-A, "RSG LOCA - B&W Loss of Coolant Accident Evaluation Model forRecirculating Steam Generator Plants" November 16, 2006 SEQUOYAH - UNIT 1 6-13 Amendment No. 52, 58, 72, 74, 117, 152, 155, 156, 171, 216, 223, 281, 300, 314 5.6.3 5.6.3.a 5.6.3.b Page 4 of 16 SL 2.1.1, "Reactor Core Safety Limits" LCO 3.1.1, SHUTDOWN MARGIN" LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower T Nominal Trip Setpoint denoted values LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" LCO 3.9.1, "Boron Concentration"A05A04Revision 0, , March 1993Revision 1, , October 1989 Revision 0, , June 1989A07EMF-2328 (P)(A), "PWR Small Break LOCA Evaluation Model," March 2001 A01ITS ITS 5.6 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)5.WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using theNOTRUMP Code"6.WCAP-10266-P-A, "The 1981 Revision of Westinghouse Evaluation Model Using BASHCODE"7.BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWRReactor Fuel" 8.BAW-10186-A, "Extended Burnup Evaluation"9.EMF-2103P-A, "Realistic Large Break LOCA Methodology for Pressurized WaterReactors" 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. 6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be providedwithin 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: Specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: 1.Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop ColdOverpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."2.Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation."3.Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel FlangeRequirements Evaluation for Sequoyah Units 1 and 2."6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto. STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (SG) Program. The report shall include: September 24, 2008 SEQUOYAH - UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117, 155, 223, 241, 258, 294, 297, 306, 314, 320 5.6.6 5.6.3.c 5.6.3.d A065.6.4 5.6.4.a 5.6.4.b 5.6.4.c Page 5 of 16 M02Revision 1, , June 2003 , June 2003Revision 2, Revision 0, , April 2003 5 6 7 INSERT 2 A07 ITS 5.6 Insert Page 6-13a INSERT 2 8.BAW-1 0241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 20059.BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 199610.BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"January 199611.BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid FuelAssemblies," August 199012.BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code," January2004 A07Page 6 of 16 A01ITS ITS 5.6 ADMINISTRATIVE CONTROLS a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and h. The effective plugging percentage for all plugging in each SG. SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted. 6.10 RECORD RETENTION (DELETED)

February 23, 2006 SEQUOYAH - UNIT 1 6-14 Amendment No. 42, 52, 58, 72, 74, 117, 148, 155, 163, 174, 178, 223, 233, 241, 258, 294, 297, 306 5.6.6.a 5.6.6.b 5.6.6.c 5.6.6.d 5.6.6.e 5.6.6.f 5.6.6.g 5.6.6.f Page 7 of 16 5.6 A01ITS ITS 5.6 TABLE 3.3-10 (Continued) ACTION STATEMENTS (Continued) ACTION 4 - With the number of channels less than the minimum channels required, initiate an alternate method of monitoring containment area radiation within 72 hours and either restore the inoperable channel(s) to OPERABLE status within 30 days, or prepare and submit a special report to the Commission pursuant to Specification 6.9.2.1 within the next 14 days that provides actions taken, cause of the inoperability, and plans and schedule for restoring the channels to OPERABLE status. ACTION 5 - NOTE: Also refer to the applicable action requirements from LCO 3.3.3.5 since it may contain more restrictive actions. a. With the number of channels on one or more steam generators less than the minimum channels required for either flow rate or valve position, restore the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the next 6 hours.

b. With the number of channels on one or more steam generators less than the minimum channels required for flow rate and valve position, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the next 6 hours.

April 11, 2005 SEQUOYAH - UNIT 1 3/4 3-57b Amendment No. 112, 149, 159, 301 5.6.5 See ITS 3.3.3 See ITS 3.3.3 Page 8 of 16 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4. STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS1/ 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. 6.9.1.5 DELETED

1/A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. April 13, 2009 SEQUOYAH - UNIT 2 6-11 Amendment No. 28, 34, 50, 64, 66, 107, 134, 165, 207, 223, 231, 271, 272, 289, 298 5.6 5.6 A02A02Page 9 of 16 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT1/ 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. 6.9.1.7 (Relocated to the ODCM.) ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

6.9.1.9 (Relocated to the ODCM or PCP.)

1/A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

August 2, 1993 SEQUOYAH - UNIT 2 6-12 Amendment No. 34, 50, 66, 107, 134, 159 5.6.1 by May 15 L01INSERT 1M015.6.2 5.6.1 Note, 5.6.2 Note Page 10 of 16in accordance with 10 CFR 50.36aA03 ITS 5.6 Insert Page 6-12 INSERT 1 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. M01Page 11 of 16 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: 1.f1(I) limits for Overtemperature Delta T Trip Setpoints and f2(I) limits for Overpower DeltaT Trip Setpoints for Specification 2.2.1.2.Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit forSpecification 3/4.1.1.3,3.Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,4.Control Bank Insertion Limits for Specification 3/4.1.3.6, 5.AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,6.Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and7.Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents: The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). 1.BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 19932.BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety AnalysisMethodology for Recirculating Steam Generator Plants," October 19893.BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for WestinghouseDesigned PWRs," June 19894.EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 20015.BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material(M5) in PWR Reactor Fuel," June 20036.BAW-10186P-A, Revision 2, "Extended Burnup Evaluation," June 20037.EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for PressurizedWater Reactors," April 2003September 26, 2012 SEQUOYAH - UNIT 2 6-13 Amendment No. 44, 50, 64, 66, 107, 134, 142, 146, 161, 206, 214, 223, 272, 289, 303, 324 5.6.3 5.6.3.a 5.6.3.b Page 12 of 16SL 2.1.1, "Reactor Core Safety Limits" LCO 3.1.1, SHUTDOWN MARGIN" LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower T Nominal Trip Setpoint denoted values LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB Limits"LCO 3.9.1, "Boron Concentration" A05A04 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued) 8.BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 20059.BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 199610.BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"January 199611.BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid FuelAssemblies," August 199012.BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code," January2004 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. 6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: Specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: 1.Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop ColdOverpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."2.Westinghouse Topical Report WCAP-15321, "Sequoyah Unit 2 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation."3.Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel FlangeRequirements Evaluation for Sequoyah Units 1 and 2."6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto. September 26, 2012 SEQUOYAH - UNIT 2 6-14 Amendment No. 44, 50, 64, 66, 107, 134, 146, 206, 214, 231, 249, 284, 303, 305, 311, 324 5.6.3.c 5.6.3.d A065.6.4 5.6.4.a 5.6.4.b 5.6.4.c M02Page 13 of 16 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT 6.9.1.16.1 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG) Program." The report shall include: a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

September 26, 2012 SEQUOYAH - UNIT 2 6-14a Amendment No. 305, 323, 3245.6.6 5.6.6.a 5.6.6.b 5.6.6.c 5.6.6.d 5.6.6.e 5.6.6.f 5.6.6.g 5.6.6.f Page 14 of 16 A01ITS ITS 5.6. ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4. 6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

July 10, 2012 SEQUOYAH - UNIT 2 6-15 Amendment No. 28, 44, 50, 64, 66, 107, 134, 146, 153, 165, 169, 206, 214, 223, 231, 249, 284, 309, 323 5.6 Page 15 of 16 A01ITS ITS 5.6. TABLE 3.3-10 (Continued) ACTION STATEMENTS (Continued) ACTION 4 - With the number of channels less than the minimum channels required, initiate an alternate method of monitoring containment area radiation within 72 hours and either restore the inoperable channel(s) to OPERABLE status within 30 days, or prepare and submit a special report to the Commission pursuant to Specification 6.9.2.1 within 14 days that provides actions taken, cause of the inoperability, and plans and schedule for restoring the channels to OPERABLE status. ACTION 5 - NOTE: Also refer to the applicable action requirements from LCO 3.3.3.5 since it may contain more restrictive actions. a. With the number of channels on one or more steam generators less than the minimum channels required for either flow rate or valve position, restore the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the next 6 hours. b. With the number of channels on one or more steam generators less than the minimum channels required for flow rate and valve position, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the next 6 hours.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 3-58b Amendment Nos. 102, 135, 149, 290 5.6.5 See ITS 3.3.3 See ITS 3.3.3. Page 16 of 16 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Sequoyah Unit 1 and Unit 2 Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 6.9.1.4 states that, annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year following initial criticality. ITS 5.6 does not include the requirement for annual reports. This changes the CTS by not including the requirements. The purpose of CTS 6.9.1.4 is to specify submittal dates of annual reports for associated activities. This change is acceptable because no activities are associated with the current Specification. This change is designated as administrative because it does not result in technical changes to the CTS. A03 CTS 6.9.1.8 requires the Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year to be submitted prior to May 1 of each year. ITS 5.6.2 requires this report, the Radioactive Effluent Release Report, covering the operation of the unit in the previous year to be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. This changes the CTS by explicitly stating the report shall be submitted in accordance with 10 CFR 50.36a. The purpose to CTS 6.9.1.8 is to provide the requirements associated with the Radioactive Effluent Release Report. 10 CFR 50.36a, "Technical Specifications on Effluents from Nuclear Power Reactors," also provides requirements for submission of a report to the Commission annually that specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents during the previous 12 months. 10 CFR 50.36a also states that the time between submissions of the reports must be no longer than 12 months. This change is acceptable because the CTS reporting requirements have not changed, ITS explicitly states that the reporting requirement of "prior to May 1 of each year," is also in accordance with 10 CFR 50.36a. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 6.9.1.14.a requires, in part, the COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). ITS 5.6.3 b Reviewers Note states, licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Sequoyah Unit 1 and Unit 2 Page 2 of 4 identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). This changes the CTS by not including the requirement of referencing Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). SQN has received prior approval by the NRC to include reference Topical Reports used to prepare the COLR (i.e., report number, title revision, date, and any supplements) in the Specification. This change is acceptable because the Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements) have been included in the Specification. This change is designated as administrative because it does not result in technical changes to the CTS. A05 CTS 6.9.1.14 contains a list of the core operating limits established and documented in the COLR. ITS 5.6.5.a includes additional core operating limits established and documented in the COLR. These are Reactor Core Safety Limits; SHUTDOWN MARGIN; Reactor Trip System (RTS) Instrumentation, - Overtemperature T and Overpower T Nominal Trip Setpoint denoted values; RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and Boron Concentration. These limits had previously been addressed in other parts of the CTS, but are being moved to the COLR in the ITS, and because of this are listed in ITS 5.6.5.a. This changes the CTS by adding core operating limits established and documented in the COLR because they are being moved there as part of changes to other parts of the CTS. Technical aspects of the changes are addressed in the Discussion of Changes for the respective individual ITS Specifications. This change is acceptable because it administratively documents changes made to other parts of the CTS and the COLR. This change is designated as administrative because it does not result in technical changes to the CTS. A06 CTS 6.9.1.14.c requires, in part the CORE OPERATING LIMITS REPORT (COLR) to be provided to the NRC document control desk with copies to the Regional Administrator and Resident Inspector. ITS 5.6.3.d requires the COLR to be provided to the NRC. This changes the CTS by removing the specifics regarding distribution of the report to the NRC. 10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS. A07 SQN Unit 1 CTS 6.9.1.14.a requires, in part that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, to be listed. TVA has received approval to change the list of approved documents used to determine the core operating limits. This changes the CTS by revising the list of approved documents to those approved in License Amendment 331 before it has been implemented at SQN Unit 1. DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Sequoyah Unit 1 and Unit 2 Page 3 of 4 This change is acceptable because this change was approved by License Amendment 331/324 [Unit 1/Unit 2] in September of 2012 by letter titled, "Sequoyah Nuclear Plant, Units 1 and 2 Issuance of Amendments to Revise the Technical Specification to allow use of Areva Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07) (TAC NOS. ME6538 and ME6539)" (ADAMS Accession No. ML12249A394). This amendment was effective as of its date of issuance, to be implemented on Unit 1 prior to startup from Unit 1 fall 2013 refueling outage and on Unit 2 prior to startup from Unit 2 fall 2012 refueling outage. SQN Unit 2 License Amendment 324 has been implemented on Unit 2 and is reflected in this license amendment request. Because the implementation of SQN Unit 1 License Amendment 331 is after the submittal of the SQN ITS conversion license amendment request, the values approved in License Amendment 331 are shown as being inserted. This change is designated as administrative because it does not result in technical changes to the CTS approved by the NRC. MORE RESTRICTIVE CHANGES M01 The second paragraph of ITS 5.6.1 includes details required to be included in the Annual Radiological Environmental Operating Report. CTS 6.9.1 does not contain this level of detail. This changes the CTS by requiring additional detail to be included in the Annual Radiological Environmental Operating Report. The purpose of the second paragraph of ITS 5.6.1 is to specify details to be included in the Annual Radiological Environmental Operating Report. This change is acceptable because the content requirements are consistent with the objectives outlined in the Offsite Dose Calculation Manual. This change is designated more restrictive because it adds new reporting requirements to the Technical Specifications. M02 CTS 6.9.1.14.c states, in part, that the CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle. ITS 5.6.3.d states, in part, that the COLR shall be provided within 30 days of issuance for each reload cycle to the NRC. This changes the CTS by eliminating the allowance to wait until entering MODE 2 for the 30 day period to begin before requiring the COLR to be submitted to the NRC. The purpose to CTS 6.9.1.14.c is to provide guidance on when the COLR is required to be submitted to the NRC. ITS 5.6.3.d provides similar guidance but requires the COLR to be submitted in less time than allowed by CTS. This change is acceptable because the ITS requirement for submission of the COLR continues to allow adequate time to process the submittal and is within the CTS requirements. This change is designated as more restrictive because less time is allowed in ITS to submit the COLR than is allowed in CTS. RELOCATED SPECIFICATIONS None DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Sequoyah Unit 1 and Unit 2 Page 4 of 4 REMOVED DETAIL CHANGES None

LESS RESTRICTIVE CHANGES L01 L01 (Category 1 -Relaxation of LCO Requirements) CTS 6.9.1.6 requires the Annual Radiological Environmental Operating Report to be submitted prior to May 1 of each year. ITS 5.6.1 requires the Annual Radiological Environmental Operating Report to be submitted by May 15 of each year. This changes the CTS by allowing additional time to submit this report each year. The purpose of the due date for submitting the Annual Radiological Environmental Operating Report is to ensure that the report is provided in a reasonable period of time to the NRC for review. This change is acceptable because the report is still required to be provided to the NRC on or before May 15 and cover the previous calendar year, report completion and submittal is clearly not necessary to assure operation in a safe manner for the interval between May 1 and May 15. Additionally, there is no requirement for the NRC to approve the report. This change is designated as less restrictive because it allows more time to prepare and submit the report to the NRC. Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Reporting Requirements 5.6 Westinghouse STS 5.6-1 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX25.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements

The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Annual Radiological Environmental Operating Report

  ----------------------------------------REVIEWER'S NOTE----------------------------------------    [ A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. ]    ------------------------------------------------------------------------------------------------------------

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.2 Radioactive Effluent Release Report ----------------------------------------REVIEWER'S NOTE---------------------------------------- [ A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ] ------------------------------------------------------------------------------------------------------------

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. 6.9.1.6 6.9 6.9.1 6.9.1 Note 6.9.1 Note 6.9.1.6 DOC M01 6.9.1.8 6.9.1.8 11DOC L01 1 Reporting Requirements 5.6 Westinghouse STS 5.6-2 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX25.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

   [ The individual specifications that address core operating limits must be referenced here. ]

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: --------------------------------REVIEWER'S NOTE---------------------------------------- Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details. ----------------------------------------------------------------------------------------------------

   [ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

   [ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]  6.9.1.14 6.9.1.14 INSERT 1 INSERT 2INSERT 3 6.9.1.14.a 6.9.1.14.b 6.9.1.14.c 6.9.1.15 6.9.1.15 1131within 30 days of2 ITS 5.6 Insert Page 5.6-2a INSERT 1 1.SL 2.1.1, "Reactor Core Safety Limits";2.LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3.LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";4.LCO 3.1.5, "Shutdown Bank Insertion Limits";5.LCO 3.1.6, "Control Bank Insertion Limits"; 6.LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(X, Y, Z))";7.LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y))";8.LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";9.LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T andOverpower T Nominal Trip Setpoint denoted values; 10.LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits"; and 11.LCO 3.9.1, "Boron Concentration."1 ITS 5.6 Insert Page 5.6-2b INSERT 2 1.BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 19932.BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety AnalysisMethodology for Recirculating Steam Generator Plants," October 19893.BAW-10163P-A, Revision 0,"Core Operating Limit Methodology for Westinghouse-DesignedPWRs," June 19894.EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 20015.BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5)in PWR Reactor Fuel," June 20036.BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003 7.EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for PressurizedWater Reactors," April 20038.BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 20059.BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 199610.BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"January 199611.BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid FuelAssemblies," August 1990 and12.BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code" January2004. INSERT 3 1.LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits"; 2.LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System"; and3.LCO 3.5.2, "ECCS - Operating".11 Reporting Requirements 5.6 Westinghouse STS 5.6-3 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX25.6 Reporting Requirements

5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: --------------------------------REVIEWER'S NOTE---------------------------------------- Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the PTLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details. ----------------------------------------------------------------------------------------------------

   [ Identify the NRC staff approval document by date.]

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

  ----------------------------------------REVIEWER'S NOTE----------------------------------------    The methodology for the calculation of the P-T limits for NRC approval should include the following provisions:

1. The methodology shall describe how the neutron fluence is calculated (reference new Regulatory Guide when issued).

2. The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR 50. The reactor vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.

3. Low Temperature Overpressure Protection (LTOP) System lift setting limits for the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR.

4. The adjusted reference temperature (ART) for each reactor beltline material shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide 1.99, Revision 2. 5. The limiting ART shall be incorporated into the calculation of the pressure and temperature limit curves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits. 6. LTOP arming temperature limit development methodology. INSERT 46.9.1.15.a 6.9.1.15.b 133within 30 days of2 ITS 5.6 Insert Page 5.6-3 INSERT 4 1. Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"; 2. Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation"; and 3. Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2." 1 Reporting Requirements 5.6 Westinghouse STS 5.6-4 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXX25.6 Reporting Requirements

5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

7. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

8. Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology. ------------------------------------------------------------------------------------------------------------

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.[3], "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. 5.6.6 [ Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. ] 5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, 6.9.1.16 6.9.1.16.a 6.9.1.16.b 6.9.1.16.c 3.3.10 ACTION 4 6 7 34146I5;8;;

Reporting Requirements 5.6 Westinghouse STS 5.6-5 Rev. 4.0 CTS SEQUOYAH UNIT 1 Amendment XXXx 5.6 Reporting Requirements

5.6.7 Steam Generator Tube Inspection Report (continued)

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged [or repaired] during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged [or repaired] to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, [h. The effective plugging percentage for all plugging [and tube repairs] in each SG, and] [i. Repair method utilized and the number of tubes repaired by each repair method.] 6.9.1.16.d 6.9.1.16.e 6.9.1.16.f 6.9.1.16.g 6.9.1.16.h 6 47;88;;; and .

Reporting Requirements 5.6 Westinghouse STS 5.6-1 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX25.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements

The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Annual Radiological Environmental Operating Report

  ----------------------------------------REVIEWER'S NOTE----------------------------------------    [ A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. ]    ------------------------------------------------------------------------------------------------------------

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.2 Radioactive Effluent Release Report ----------------------------------------REVIEWER'S NOTE---------------------------------------- [ A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ] ------------------------------------------------------------------------------------------------------------

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. 6.9.1.6 6.9 6.9.1 6.9.1 Note 6.9.1 Note 6.9.1.6 DOC M01 6.9.1.8 6.9.1.8 11DOC L01 1 Reporting Requirements 5.6 Westinghouse STS 5.6-2 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX25.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

   [ The individual specifications that address core operating limits must be referenced here. ]

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: --------------------------------REVIEWER'S NOTE---------------------------------------- Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details. ----------------------------------------------------------------------------------------------------

   [ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

   [ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]  6.9.1.14 6.9.1.14 INSERT 1 INSERT 2INSERT 3 6.9.1.14.a 6.9.1.14.b 6.9.1.14.c 6.9.1.15 6.9.1.15 1131within 30 days of2 ITS 5.6 Insert Page 5.6-2a INSERT 1 1.SL 2.1.1, "Reactor Core Safety Limits";2.LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3.LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";4.LCO 3.1.5, "Shutdown Bank Insertion Limits";5.LCO 3.1.6, "Control Bank Insertion Limits"; 6.LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(X, Y, Z))";7.LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y))";8.LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";9.LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T andOverpower T Nominal Trip Setpoint denoted values; 10.LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits"; and 11.LCO 3.9.1, "Boron Concentration."1 ITS 5.6 Insert Page 5.6-2b INSERT 2 1.BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 19932.BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety AnalysisMethodology for Recirculating Steam Generator Plants," October 19893.BAW-10163P-A, Revision 0,"Core Operating Limit Methodology for Westinghouse-DesignedPWRs," June 19894.EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 20015.BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5)in PWR Reactor Fuel," June 20036.BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003 7.EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for PressurizedWater Reactors," April 20038.BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 20059.BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 199610.BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"January 199611.BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid FuelAssemblies," August 1990 and12.BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code" January2004. INSERT 3 1.LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits"; 2.LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System"; and3.LCO 3.5.2, "ECCS - Operating".11 Reporting Requirements 5.6 Westinghouse STS5.6-3Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX25.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued) b.The analytical methods used to determine the RCS pressure andtemperature limits shall be those previously reviewed and approved by theNRC, specifically those described in the following documents: --------------------------------REVIEWER'S NOTE----------------------------------------Licensees that have received prior NRC approval to relocate Topical Reportrevision numbers and dates to licensee control need only list the number and title of the Topical Report, and the PTLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details. ---------------------------------------------------------------------------------------------------- [ Identify the NRC staff approval document by date.] c.The PTLR shall be provided to the NRC upon issuance for each reactorvessel fluence period and for any revision or supplement thereto. ----------------------------------------REVIEWER'S NOTE---------------------------------------- The methodology for the calculation of the P-T limits for NRC approval should include the following provisions: 1.The methodology shall describe how the neutron fluence is calculated(reference new Regulatory Guide when issued). 2.The Reactor Vessel Material Surveillance Program shall comply withAppendix H to 10 CFR 50. The reactor vessel material irradiationsurveillance specimen removal schedule shall be provided, along with howthe specimen examinations shall be used to update the PTLR curves. 3.Low Temperature Overpressure Protection (LTOP) System lift setting limitsfor the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR. 4.The adjusted reference temperature (ART) for each reactor beltline materialshall be calculated, accounting for radiation embrittlement, in accordancewith Regulatory Guide 1.99, Revision 2. 5.The limiting ART shall be incorporated into the calculation of the pressureand temperature limit curves in accordance with NUREG-0800 StandardReview Plan 5.3.2, Pressure-Temperature Limits. 6.LTOP arming temperature limit development methodology.INSERT 46.9.1.15.a 6.9.1.15.b 133within 30 days of2 ITS 5.6 Insert Page 5.6-3 INSERT 4 1.Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop ColdOverpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves";2.Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation"; and3.Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel FlangeRequirements Evaluation for Sequoyah Units 1 and 2."1 Reporting Requirements 5.6 Westinghouse STS 5.6-4 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXX25.6 Reporting Requirements

5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

7. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

8. Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology. ------------------------------------------------------------------------------------------------------------

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.[3], "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. 5.6.6 [ Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. ] 5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, 6.9.1.16 6.9.1.16.a 6.9.1.16.b 6.9.1.16.c 3.3.10 ACTION 4 6 7 34146I5;8;;

Reporting Requirements 5.6 Westinghouse STS 5.6-5 Rev. 4.0 CTS SEQUOYAH UNIT 2 Amendment XXXx 5.6 Reporting Requirements

5.6.7 Steam Generator Tube Inspection Report (continued)

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged [or repaired] during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged [or repaired] to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, [h. The effective plugging percentage for all plugging [and tube repairs] in each SG, and] [i. Repair method utilized and the number of tubes repaired by each repair method.] 6.9.1.16.d 6.9.1.16.e 6.9.1.16.f 6.9.1.16.g 6.9.1.16.h 6 47;88;;; and .

JUSTIFICATION FOR DEVIATIONS ITS 5.6, STEAM GENERATOR TUBE INSPECTION REPORT Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. 4. ISTS 5.6.6 provides requirements for the Tendon Surveillance Report. The Containment design at SQN does not include pre-stressed concrete tendons. Therefore, this report is not included in the SQN ITS, consistent with the current licensing basis. Subsequent Specifications are renumbered as a result of this deletion. 5. Changes made to reflect those changes made to ITS 3.3.3, "Post Accident Monitoring (PAM) Instrumentation."

6. Changes made to reflect those changes made to ITS 5.5.7, "Steam Generator (SG) Program."

7. Sequoyah Unit 1 and 2 are not licensed for repair of SG tubes, so the bracketed allowance has been deleted.

8. These punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications NEI 01-03, Section 5.1.3.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ATTACHMENT 7 ITS 5.7, HIGH RADIATION AREA Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM(DELETED) 6.12 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20: 6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment. b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. d. Each individual or group entering such an area shall possess: 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance. e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. February 11, 2003 SEQUOYAH - UNIT 1 6-15 Amendment No. 42, 58, 74, 148, 152, 174, 178, 212, 233, 266, 281 that includes specification of radiation dose rates in the immediate work area(s) A025.7 5.7.1 5.7.1.a 5.7.1.b 5.7.1.c 5.7.1.d 5.7.1.d.1 5.7.1.d.2 5.7.1.d.3 5.7.1.d.4 5.7.1.e 5.7.1.d.4.(i) 5.7.1.d.4.(ii) Page 1 of 6 A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition: 1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee. 2. Doors and gates shall remain locked except when needed for personnel or equipment access. b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. d. Each individual or group entering such an area shall possess: 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area. February 11, 2003 SEQUOYAH - UNIT 1 6-15a Amendment No. 281 during periods ofentry or exit that includes specifications of dose rates in the immediate work area(s) A02or 5.7.2 5.7.2.a 5.7.2.a.1 5.7.2.a.2 5.7.2.b 5.7.2.c 5.7.2.d 5.7.2.d.1 5.7.2.d.2 5.7.2.d.3 5.7.2.d.3.(i) 5.7.2.d.3.(ii) 5.7.2.d.4 Page 2 of 6 s and A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

February 11, 2003 SEQUOYAH - UNIT 1 6-15b Amendment No. 281 5.7.2.e 5.7.2.f Page 3 of 6 A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM (DELETED 6.12 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20: 6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment. b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. d. Each individual or group entering such an area shall possess: 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance. e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. February 11, 2003 SEQUOYAH - UNIT 2 6-16 Amendment No. 34, 50, 66, 134, 142, 165, 169, 202, 223, 257, 272 Page 4 of 6 that includes specification of radiation dose rates in the immediate work area(s) A025.7 5.7.1 5.7.1.a 5.7.1.b 5.7.1.c 5.7.1.d 5.7.1.d.1 5.7.1.d.2 5.7.1.d.3 5.7.1.d.4 5.7.1.e 5.7.1.d.4.(i) 5.7.1.d.4.(ii) A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition: 1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee. 2. Doors and gates shall remain locked except when needed for personnel or equipment access. b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. d. Each individual or group entering such an area shall possess: 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area. February 11, 2003 SEQUOYAH - UNIT 2 6-16a Amendment No. 272 during periods ofentry or exit that includes specifications of dose rates in the immediate work area(s) A02or Page 5 of 6 5.7.2 5.7.2.a 5.7.2.a.1 5.7.2.a.2 5.7.2.b 5.7.2.c 5.7.2.d 5.7.2.d.1 5.7.2.d.2 5.7.2.d.3 5.7.2.d.3.(i) 5.7.2.d.3.(ii) 5.7.2.d.4 s and A01ITS ITS 5.7 ADMINISTRATIVE CONTROLS e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

February 11, 2003 SEQUOYAH - UNIT 2 6-16b Amendment No. 272 Page 6 of 6 5.7.2.e 5.7.2.f DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA Sequoyah Unit 1 and Unit 2 Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A02 CTS 6.12.1.b and CTS 6.12.2.b state, in part, access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent and associated radiation survey. ITS 5.7.1.b and ITS 5.7.2.b state, in part, that access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area (s). This changes the CTS by specifying the document equivalent to the RWP shall include specification of radiation dose rates in the immediate work area(s). The purpose of CTS 6.12.1.b and CTS 6.12.2.b is to specify the controls needed to access high radiation areas. This change is acceptable because the additional wording that the RWP equivalent includes a specification of radiation dose rates in the immediate work area(s) clarifies the requirements of an RWP. This is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) High Radiation Area 5.7 Westinghouse STS 5.7-1 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1 5.0 ADMINISTRATIVE CONTROLS

[ 5.7 High Radiation Area ] As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures. c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area, or 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 6.12 6.12.1 6.12.1.a 6.12.1.b 6.12.1.c 6.12.1.d 6.12.1.d.1 6.12.1.d.2 6.12.1.d.3 6.12.1.d.4 6.12.1.d.4.(i) 1 High Radiation Area 5.7 Westinghouse STS 5.7-2 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1 5.7 High Radiation Area

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designees, and

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit. b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. 6.12.1.d.4.(ii) 6.12.1.e 6.12.2 6.12.2.a 6.12.2.a.1 6.12.2.a.2 6.12.2.b 6.12.2.c 6.12.1 manager2or 2 High Radiation Area 5.7 Westinghouse STS 5.7-3 Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 1 5.7 High Radiation Area

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

d. Each individual group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displaces radiation dose rates in the area. e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. 6.12.2 6.12.2.d 6.12.2.d.1 6.12.2.d.2 6.12.2.d.3 6.12.2.d.3.(i) 6.12.2.d.3.(ii) 6.12.2.d.4 6.12.2.e displays 3or2or 2 High Radiation Area 5.7 Westinghouse STS5.7-4Rev. 4.0 CTS 2SEQUOYAH UNIT 1 Amendment XXX 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) f.Such individual areas that are within a larger area where no enclosureexists for the purpose of locking and where no enclosure can reasonably beconstructed around the individual area need not be controlled by a lockeddoor or gate, nor continuously guarded, but shall be barricaded,conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.6.12.2 6.12.2.f High Radiation Area 5.7 Westinghouse STS5.7-1Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2 5.0 ADMINISTRATIVE CONTROLS [ 5.7 High Radiation Area ] As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20: 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation a.Each entryway to such an area shall be barricaded and conspicuouslyposted as a high radiation area. Such barricades may be opened asnecessary to permit entry or exit of personnel or equipment.b.Access to, and activities in, each such area shall be controlled by means ofRadiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.c.Individuals qualified in radiation protection procedures and personnelcontinuously escorted by such individuals may be exempted from therequirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.d.Each individual or group entering such an area shall possess:1.A radiation monitoring device that continuously displays radiation doserates in the area, or2.A radiation monitoring device that continuously integrates the radiationdose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or3.A radiation monitoring device that continuously transmits dose rate andcumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or4.A self-reading dosimeter (e.g., pocket ionization chamber or electronicdosimeter) and,(i) Be under the surveillance, as specified in the RWP or equivalent,while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 6.12 6.12.1 6.12.1.a 6.12.1.b 6.12.1.c 6.12.1.d 6.12.1.d.1 6.12.1.d.2 6.12.1.d.3 6.12.1.d.4 6.12.1.d.4.(i) 1 High Radiation Area 5.7 Westinghouse STS5.7-2Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance. e.Except for individuals qualified in radiation protection procedures, orpersonnel continuously escorted by such individuals, entry into such areasshall be made only after dose rates in the area have been determined andentry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation a.Each entryway to such an area shall be conspicuously posted as a highradiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:1.All such door and gate keys shall be maintained under theadministrative control of the shift supervisor, radiation protectionmanager, or his or her designees, and 2.Doors and gates shall remain locked except during periods ofpersonnel or equipment entry or exit.b.Access to, and activities in, each such area shall be controlled by means ofan RWP or equivalent that includes specification of radiation dose rates inthe immediate work area(s) and other appropriate radiation protection equipment and measures.c.Individuals qualified in radiation protection procedures may be exemptedfrom the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plantradiation protection procedures for entry to, exit from, and work in such areas.6.12.1.d.4.(ii) 6.12.1.e 6.12.2 6.12.2.a 6.12.2.a.1 6.12.2.a.2 6.12.2.b 6.12.2.c 6.12.1 manager2or 2 High Radiation Area 5.7 Westinghouse STS5.7-3Rev. 4.0 CTS 2Amendment XXX SEQUOYAH UNIT 2 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) d.Each individual group entering such an area shall possess:1.A radiation monitoring device that continuously integrates the radiationrates in the area and alarms when the device's dose alarm setpoint isreached, with an appropriate alarm setpoint, or2.A radiation monitoring device that continuously transmits dose rate andcumulative dose information to a remote receiver monitored byradiation protection personnel responsible for controlling personnelradiation exposure within the area with the means to communicate with and control every individual in the area, or3.A self-reading dosimeter (e.g., pocket ionization chamber or electronicdosimeter) and,(i) Be under surveillance, as specified in the RWP or equivalent,while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. 4.In those cases where options (2) and (3), above, are impractical ordetermined to be inconsistent with the "As Low As is ReasonablyAchievable" principle, a radiation monitoring device that continuouslydisplaces radiation dose rates in the area.e.Except for individuals qualified in radiation protection procedures, orpersonnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. 6.12.2 6.12.2.d 6.12.2.d.1 6.12.2.d.2 6.12.2.d.3 6.12.2.d.3.(i) 6.12.2.d.3.(ii) 6.12.2.d.4 6.12.2.e displays 3or2or 2 High Radiation Area 5.7 Westinghouse STS5.7-4Rev. 4.0 CTS 2SEQUOYAH UNIT 2 Amendment XXX 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) f.Such individual areas that are within a larger area where no enclosureexists for the purpose of locking and where no enclosure can reasonably beconstructed around the individual area need not be controlled by a lockeddoor or gate, nor continuously guarded, but shall be barricaded,conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.6.12.2 6.12.2.f JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA Sequoyah Unit 1 and Unit 2 Page 1 of 1 1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Typographical/grammatical error corrected.

Specific No Significant Hazards Considerations (NSHCs) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA Sequoyah Unit 1 and 2 Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification. ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 Risk Informed Evaluation of Extensions to Containment Isolation Valve Completion Times (WCAP-15791) Risk Informed Evaluation of Extensions to Containment Isolation Valve Completion Times (WCAP-15791) 1.0 Purpose This analysis considers the Sequoyah (SQN) as-built, as-operated plant to ascertain acceptability of applying NRC endorsed topical report (TR) WCAP-15791-P-A Revision 2 "Risk-Informed Evaluation of Extensions to Containment Isolation Valve Completion Times."

The benefit of this proposed change to the Technical Specifications is that completion time (CT) extensions will provide the operator flexibility by increasing the time to perform on-line CIV testing, maintenance or repair. Currently CIV completion time is limited to four-hours for all CIVs. 2.0 References and Acronyms 2.1 References

1. NUREG/CR-5496, "CCF Parameter Estimation 2007" 2. Regulatory Guide 1.174, Rev. 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" 3. Regulatory Guide 1.177, Rev. 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" 4. Regulatory Guide 1.200, Rev. 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" 5. WCAP-15791-P-A, Rev. 2 "Risk-Informed Evaluation of Extensions to Containment Isolation Valve Completion Times" 6. NRC Accession Number 10027058, Generic Issue (GI-199), "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," August 2010 7. MDN-000-000-2010-0200, Rev. 1 - SQN PRA - "Summary Notebook" 8. MDN-000-000-2010-0202, Rev. 1 - SQN PRA - "Data Analysis" 9. MDN-000-000-2010-0203, Rev. 1 - SQN PRA - "Internal Flooding Analysis"

10. MDN-000-000-2010-0208, Rev. 2 - SQN PRA - "Quantification Notebook"

11. 1-SI-SXV-000-201.0, Rev. 017 - Surveillance Instruction - "Full Stroking of Category 'A' and 'B' Valves During Operation" 12. N2-88-400, Rev. 15 System Description - "Containment Isolation"

13. MDQ-000088-2013-000072 Rev. 0 "Risk-Informed Evaluation of Extension to Containment Isolation Valve Completion Times" 14. LTR-RAM-II-11-010, "R.G. 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Sequoyah Nuclear Plant Probabilistic Risk Assessment," March 2011

2.2 Acronyms The following acronyms are used in this analysis: AOT - Allowed Outage Time AOV - Air Operated Valves ASME - American Society of Mechanical Engineers CAFTA - Computer Aided Fault Tree Analysis CCF - Common Cause Failure CDE - Cause Determination and Evaluation CDF - Core Damage Frequency CDFSEIS - Core Damage Frequency due to Seismic Event CDFSGTR- Core Damage Frequency due to Steam Generator Tube Rupture CDFT - Total Core Damage Frequency (Internal & External Events) CIV - Containment Isolation Valve CKV - Check Valve CRMP - Configuration Risk Management Program CT - Completion Time FTC - Fail-To-Close ICCDP - Incremental Conditional Core Damage Probability ICLERP - Incremental Conditional Large Early Release Probability IPE - Individual Plant Examination IPEEE - Individual Plant Examination of External Events ISLOCA - Interfacing System Loss of Coolant Accident LCO - Limiting Condition for Operation LERF - Large Early Release Frequency LLRT - Local Leak Rate Testing MOV - Motor Operated Valves MGL - Multiple Greek Letter NRC - Nuclear Regulatory Commission NRR - Nuclear Reactor Regulation PBRANDOM - Pipe Break - Random PBSEIS - Pipe Break - Seismic PRA - Probabilistic Risk Assessment PWROG - Pressurized Water Reactor Owner's Group RCS - Reactor Coolant System RG - Regulatory Guide SE - Safety Evaluation SG - Steam Generator SGTR - Steam Generator Tube Rupture SI - Surveillance Instruction SOV - Solenoid Operated Valve SQN - Sequoyah Nuclear Power Plant SRP - Standard Review Plan SRV - Safety Relief Valve SSE - Safe Shutdown Earthquake TR - Topical Report TS - Technical Specifications WCAP - Westinghouse Commercial Atomic Power WOG - Westinghouse Owner's Group XO - Spurious (Transfers) Open 3.0 Assumptions and Analysis Basis 1. Before maintenance or repair is started on a containment isolation valve (CIV), it is assumed the other CIVs within the penetration are verified by Operations to be in their proper position(s).1 2. It is assumed that manually operated vent or drain valves located between the CIVs are verified by Operations to be in their closed position similar to assumption 1, as well as other normally closed manually operated valves connected to the penetration. 3. Manually operated vent or drains valves, if opened for LLRT (Local Leak Rate Testing), etc., are assumed to have a completion time based on the most restrictive CIV for the penetration. This is because the vent/drain lines are less than or equal to the CIV diameter, therefore, their CT is bounded by the larger valves in the associated penetration. 4. It is assumed that containment isolation valves that are locked closed have been verified closed (and locked to prevent inadvertent opening) by operations and therefore excluded from the analysis as potential to spuriously transfer open. 5. For this analysis, it is conservatively assumed that valves that are periodically opened/closed (e.g., containment purge valves) are normally opened. 6. If a seismic event greater than a SSE (Safe-Shutdown Earthquake) were to occur, it is assumed that all non-seismically qualified piping will fail, i.e., a probability of 1.0 is given. Sections of pipe between the containment isolation valve and the containment wall which is part of the break exclusion zone are excluded.2 7. Containment isolation valves are tested quarterly [Ref 11], additionally it is assumed there is one miscellaneous CIV actuation per year making a total of five actuations per year.3 8. Regardless of the completion time (CT) it is conservatively assumed the maintenance activity requires the entire length of the CT and is completed within that time. 9. For penetration configurations that have an "extra valve" i.e., not a CIV; the probability assigned for those non-CIV valves being in maintenance remains constant for all CTs. These valves are mainly recognized in the Reactor Coolant System (RCS) penetrations.4 The assumption is that the CT on all non-CIV valves modeled in this analysis is 72 hours. 10. Only one valve within a single containment penetration can be in maintenance at a time.

11. Maintenance on a valve can be conducted in one of two ways: a) the valve is intact and capable of maintaining its pressure boundary function, or b) the valve is not intact and is not capable of maintaining its pressure boundary function. 12. When there are two or more valves of the same valve type in the same position (opened or closed) within a penetration, common cause failures (CCF) are included in the ICLERP and LERF calculations.5 1 This assumption eliminates the need to include the probability that the operable valves were mispositioned or transferred to the wrong position since they were last checked. This approach is consistent with WCAP-15791. 2 The piping in the break exclusion zone is more robust that the piping outside the zone. Therefore, consistent with the approach taken in WCAP 15791 it is assumed that the probability of this piping failing randomly or due to a seismic event is much lower than the piping outside the exclusion zone and is of no consequence to this analysis. 3 Same approach as taken in the WCAP. 4 The extra valve(s) provides an additional capability for the operator to isolate a penetration. 5 For cases whereby one CIV is out-of-service for repair, the second CIV of the same valve type has a dependent failure probability involving the common-cause beta factor. For cases where there are three CIVs of the same type, the dependent failure probability involves the gamma factor. The Multiple Greek Letter (MGL) common- cause methodology is used in this analysis. Note - different valve manufacturers is irrelevant to this analysis.

13. It is assumed that interfacing system LOCAs (ISLOCAs) result in core damage. 14. Similar to the Lead Plant, SQN does not have a full scope PRA; therefore, the internal and external at-power CDF is conservatively assumed to be <1.0E-4/yr. This value represents the upper bound for the total-at-power internal and external events CDF based on the acceptance guidelines in Regulatory Guide 1.174, Section 2.2.4, that indicates the plant total CDF should be less than 1.0E-04/yr if changes to a plant's licensing bases are made that can result in a small increase in plant risk. 15. For all standby systems connected to the reactor coolant system (RCS):

the system is considered "closed" inside containment and not actively connected to the RCS if there is a closed valve between the RCS and the inside containment CIV.
the system is considered "closed" outside of containment if there is an extra closed valve before the outside containment CIV, and if the piping from the RCS and the extra closed valve outside containment is qualified for high RCS pressures. 16. For systems whereby closed both inside and outside containment, the probability of a non-seismic and/or ISLOCA CDF release is extremely small and therefore excluded from the analysis due to the large number of normally closed valves available to isolate the penetration. Additionally, the likelihood of a random pipe break occurring inside and outside containment simultaneously, causing both systems inside and outside of containment to open is very small and also excluded from the analysis. Note - ALL piping between the RCS and the extra closed valve outside of containment must be qualified for high RCS pressures. 17. For RCS connections, during seismic and random pipe breaks, it is assumed that the piping fails between the CIV and the extra valve. The portion of piping between the CIV and the containment wall is part of the break exclusion zone, and therefore is assumed to remain intact while non-qualified pipe is assumed to fail. This eliminates crediting the extra valves to isolate the penetration. 18. Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of the plant charging systems, and therefore, are not considered small LOCAs or potential containment bypass pathways. 19. For all RCS connections, in which there are two valves of the same type (usually check valves), in series inside containment, before the RCS, common cause failure does not apply because the valves are operating under different conditions. The valve closer to the RCS is subject to a higher pressure than the downstream valve. 20. For the probability of an ISLOCA release portion of the LERF calculations, when there is a normally open valve in the penetration, the open valve is not credited in the calculation. When assessing ISLOCA, the initiating event is the frequency of the closed valves within the path of release spuriously transferring open or rupturing, thus creating a flow path directly from the RCS to the outside atmosphere. 21. For all RCS connections that are normally operating, the probability of an ISLCOA release is not considered because the valves are already open and flow is occurring. 4.0 Regulatory Acceptability and Discussion 4.1 Regulatory Acceptability The Office of Nuclear Reactor Regulation (NRR) issued a safety evaluation (SE) on WCAP-15791-P, Revision 2. The SE considers Technical Specification (TS) Limiting Conditions for Operation (LCO) that state the primary containment isolation valves (CIVs) must be operable for a given reactor mode of operation.

The SE concluded: 1. The TR (topical report) provides guidance including generic and plant-specific analyses to assist licensees in evaluating changes to CIV completions times (CTs). 2. The guidance is complementary to NRC Staff guidance provided in -

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" [Ref. 2]
Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications" [Ref.3]

Therefore, NRR stated the TR provides an acceptable basis to evaluate the proposed CIV CTs when used in conjunction with the RGs. Furthermore, with respect to the acceptance criteria associated with RG 1.177, the TR addresses Tiers 1 (Probabilistic Risk Assessment Capability and Insights) and 2 (Avoidance of Risk-Significant Plant Configurations). Tier 3 (Risk-Informed Configuration Risk Management) is not addressed by the TR and must be addressed in the plant-specific application.

4.2 Discussion The TR provides a risk-informed justification for extending CIV CTs from 4-hours up to 168-hours for Westinghouse pressurized water reactors. For CIVs that do not demonstrate acceptable results for 168 hours, shorter CTs were evaluated in the report. A deterministic approach was used to determine the minimum containment hole size that would result in a large release from the containment atmosphere. These flow-paths are automatically given the 168-hours CT. All other penetrations were evaluated in the report using a PRA evaluation to verify what CT (i.e., less than 168-hours) is justified. 4.3 Regulatory Criteria 4.3.1 Standard Review Plan (SRP) 19.2 In accordance with SRP 19.2 a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles: 1. The proposed change meets current regulations, unless it explicitly relates to a requested exemption or rule change. 2. The proposed change is consistent with the defense-in-depth philosophy. 3. The proposed change maintains sufficient safety margins.

4. When proposed changes increase risk (i.e., core damage frequency (CDF) or large early release frequency (LERF)), the increase(s) should be small and consistent with the intent of the Commission's Safety Goal Policy Statement. 5. The impact of the proposed change should be monitored using performance measurement strategies.

4.3.2 Regulatory Guide 1.177 RG 1.177 provides an approach for plant-specific, risk-informed decision making for changes to the technical specifications. A three-tiered approach for evaluation of the risk associated with the proposed TS change follows: Tier 1 - an evaluation of the plant-specific risk associated with the proposed TS change, as shown by the change in:

Change in the Core Damage Frequency (CDF)
Incremental Conditional Core Damage Probability (ICCDP)
Change in Large Early Release Frequency (LERF)
Incremental Conditional Large Early Release Probability (ICLERP) Tier 2 - identifies and evaluates, with respect to defense-in-depth, any potential risk-significant plant equipment outage configurations associated with the proposed change. Tier 3 - provides for the establishment of an overall configuration risk management program (CRMP) and confirmation that its insights are incorporated into the decision making process before taking equipment out-of-service prior to or during the CT.

4.4 Generic Assessment (From WCAP-15791) WCAP-15791, "Risk-Informed Evaluation of Extensions to Containment Isolation Valve Completion Times" documents the generic analysis performed by Westinghouse for the PWR Owner's Group (PWROG). The generic assessment of impact on risk is documented in section 8 of the WCAP. The penetration configurations used in the analysis were developed to be as generic as possible. Some of the configurations may not exist within all plants, and/or some of the maintenance situations may or may not be viable for all plants. For plant-specific implementation of the generic analysis, the expectation of the WCAP is that all utilities determine the applicability of the CT in practice. In the generic analysis, Table 8-1 provides the list of input parameters. The majority of the inputs used were obtained from PRA data; however, to make the analysis as generic as possible, the most limiting (e.g., highest failure rate) values were chosen from a plant-to-plant comparison. The approach used both deterministic and probabilistic inputs. A deterministic approach was used to determine the minimum containment hole size (>2 inches) that will result in a large release from the containment atmosphere. All other penetrations are evaluated on a probabilistic basis to demonstrate if a CT of 7-days is acceptable or to determine an appropriate lesser CT. 4.5 Methodology The lead Plant followed the generic analysis, as does the SQN specific analysis. The implementation procedure followed consisted of five steps. For penetration configurations whereby completion times were less than the maximum allowable value of 168-hrs, a plant specific analysis is performed. Not all SQN penetration configurations were addressed by the WCAP of which a plant specific analysis is performed. 4.5.1 Step 1 Containment Penetration Data Collection6 This data was provided to the PRA Engineer and documented in calculation MDQ-000088-2013-000072. [Ref. 13] 4.5.2 Step 2 Confirmation of Analysis Input Parameters The generic analysis documented in WCAP-15791 used a set of input parameters that were obtained from industry PRA data. To make the analysis generic, the most limiting values were chosen. A review is performed to confirm that parameters used in the SQN PRA are bounded by the generic analysis inputs. For those that are not bounded the calculation is re-performed using the SQN parameter. 4.5.3 Step 3 Grouping Penetrations are grouped based on whether it is an open or closed system and the following attributes:

Connected to Containment Atmosphere (Class I)
Connected to the Reactor Coolant System (Class II)
Connected to the Steam Generators (Class III) 4.5.4 Step 4 Identification of Small Lines Small lines is a characterization based on the size of a hole in the primary containment that is the threshold for accident condition radionuclide large release to the environment. Note that the "Small Lines" characterization is applicable only to those penetrations connected to the containment atmosphere, i.e., Class I. 4.5.5 Step 5 Generic Match For those penetrations that did not screen from further consideration in step 4 (i.e., 168-hr CT), a comparison of the generic penetrations/flow paths listed in sections 8.2.2 through 8.2.4 of the WCAP is made. 4.5.6 Guidelines The ICLERP and/or LERF (depending on which is more limiting) was recalculated using SQN specific parameters for CIVs with CTs less than 168-hours. The inputs were used in the appropriate ICLERP and LERF equation based on the penetration Class and Group. Similar to the lead plant (Wolf Creek) analysis, two guidelines are to be followed:
For penetrations having one normally open CIV - when more than one valve type is present, use the CT for the normally open valve. All valves in the penetration will be represented by this valve type. 6 Steps 1, 3 and 4 are documented in Calculation MDQ-000088-2013-000072. [Ref. 13]
For penetrations that have more than one normally open CIV - use the CT for the normally open valve with the highest probability of failing-to-close. All valves in the penetration will be represented by this valve type.

4.6 SQN Inputs / Specific and Generic 4.6.1 Discussion The analysis involved replacing generic parameters with SQN specific parameters or updated industry data, and recalculating the probabilistic evaluation. The reason for this analysis is to determine which CIVs could be justified for longer CT relaxations in addition to those justified under the generic analysis. The approach taken in the generic analysis was conservative, and therefore, applicable to all Westinghouse Owner's Group (WOG) plants, including SQN. Where appropriate, the plant-specific analysis removes over-conservatisms. The SQN CIVs that were unable to meet the full 168 hour CT extension under the generic analysis are identified in calculation MDQ-000088-2013-000072 [Ref. 13]. The methodology, terminology, basis and assumptions that were applicable in the generic analysis are all applicable to the SQN specific analysis. The only difference is that the SQN input parameters were used in combination with generic parameters. The analyses in WCAP sections 8.2 and 8.3 were repeated using the SQN specific-parameters to calculate ICLERP and LERF. For penetration configurations that differ from the WCAP, equations were developed. The purpose of this analysis is to determine CIVs that can be justified for longer completion time (CT) relaxations (>4-hrs <168-hrs) in addition to those justified under the generic analysis. Total CDF Similar to the Lead Plant, SQN does not have a full scope PRA; therefore, the internal and external at-power CDF is conservatively assumed to be 1.0E-4/yr. This value represents the upper bound for the total-at-power internal and external events CDF based on the acceptance guidelines in Regulatory Guide 1.174, Section 2.2.4, that indicates the plant total CDF should be less than 1.0E-04/yr if changes to a plant's licensing bases are made that can result in a small increase in plant risk. Seismic SQN does not have a seismic PRA; therefore, the results documented in Table D-1, "Seismic Core-Damage Frequencies Using 2008 USGS Seismic Hazard Curves" in Generic Issue-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants" was used.[Ref 6] For SQN Units 1 & 2, the limiting frequency (5.1E-05/yr) was based on the Weakest Link Model using PGA (peak ground acceleration). Containment Isolation Valve Treatment For all normally-closed valves, the probability of the valve spuriously transferring open is considered. For all normally-open valves, the valves have the probability of a) failing-to-close, and b) spuriously transferring open after it has closed.

The limiting (bounding) valve in the SQN PRA for a spurious open or transfer open are the Safety Relief Valves. The SRVs provide for a steam release with a distribution rate of 2.12E-7/hr. Therefore, to determine the bounding probability that a valve will spuriously transfer open follows: Ptopre = 2.12E-07/hr

4 hr CT = 8.48E-07 Pto = 2.12E-07/hr
168 hrs CT = 3.56E-05 The SQN PRA does not explicitly model SOVs, therefore, SOV-FTC (9.54E-04) is used. Taken from NUREG/CR-6928.

The CIV corrective maintenance frequency (m) is derived from the highest valve failure rate used in this analysis (SOV FTC)7 which is approximately 9.54E-04 per demand, meaning the component is expected to approximately fail every 1000 actuations [(1)/(9.54E-4) ~ 1000]. Each CIV is tested quarterly [Ref 6], and in addition it is assumed that there is one8 miscellaneous CIV actuations per year. Therefore, giving a total of five CIV actuations per year. Dividing 1000 actuations by five actuations per year yields an approximate 200 year period per failure, or a corrective maintenance frequency of 0.005 per year. Therefore, the probability that a CIV is unavailable due to maintenance during a CT of 4 and 168 hours is calculated as follows: Current CT 4 hrs Pm1 = [(4-hrs /8760-hrs/yr) * (0.005)] = 2.28E-06 Extra Valve Assumed CT 72-hrs PmE = [(72-hrs/8760-hr/yr)*(0.005)] = 4.11E-05 Extended CT 168 hrs Pm2 = [(168-hrs / 8760-hrs/yr) * (0.005)] = 9.59E-05 Containment Hole Size [Ref 5 Section 8.3] Penetration flow paths connected to the containment atmosphere (this excludes RCS and SG connections) that have piping diameters smaller than a minimum value are an insufficient size to result in a large release. These penetrations automatically default to the 168-hour CT. Based on discussion with the NRC, the WOG applies a greater than 2-inch containment hole size for a large release. 4.6.2 Confirmation of Analysis Input Parameters Table 4-1 Core Damage Frequencies Input ParameterGeneric AnalysisSQN-1 SQN-2 Total Core Damage Frequency/yr CDFT 1.00E-04 1.00E-04[Note 1] 1.00E-04[Note 1] Core Damage Frequency Due to Seismic Event/yr CDFSEIS 4.41E-055.1E-05[Note 2] 5.1E-05[Note 2] 7 For SOV CIVs the generic values from the WCAP are used. SRV - Water relief are also treated independently from the generic analysis. 8 The Lead Plant analysis assumed only one additional actuation per year for a total of five. The SQN analysis applies the same assumption. Table 4-1 Core Damage Frequencies Input ParameterGeneric AnalysisSQN-1 SQN-2 Core Damage Frequency/yr Due to Steam Generator Tube Rupture CDFSGTR 9.44E-061.75E-08[Ref 7 Tbl 6] 1.78E-08[Ref 7 Tbl 6] Note 1: The SQN Internal Events + Internal Flooding Rev. 6 model quantifies a CDF of 1.59E-05 and 1.48E-05 [Ref 1] for Units 1 and 2, respectively. Similar to the generic analysis and lead plant analysis SQN does not have a full scope PRA; therefore, a generic value of 1.00E-04/yr is used in the analysis to represent the total CDF from internal and external events. Note 2: The generic core damage frequency due to a seismic event (CDFSEIS) per year was obtained from the results of GI-199. [Ref 6] The seismic frequency used in the plant specific analysis is greater than that used in the generic analysis. Therefore, generic calculations that required the large release due to seismic CDF calculation are recalculated for SQN to ascertain the CT. 4.6.3 Valve Failure Probabilities, Pftc (Per Demand) Table 4 Valve Fail-To-Close and Fail-To-Reseat Probabilities Valve Type Failure Mode Parameter Generic Analysis SQN[Note 1] AOV[Note 2] Fail-To-Close AOVftc-aov1.81E-02 5.76E-04CKV Fail-To-Close CKVftc-ckv3.44E-02 1.04E-04 MOV Fail-To-Close MOVftc-mov1.09E-02 2.77E-04 SOV Fail-To-Close SOVftc-Sov1.81E-02 9.54E-04 SRV - Steam[Note 3] Fail-To-Reseat SRVftc-srvs 2.50E-02 6.76E-05 SRV - Water Fail-To-Reseat SRVftc-srvw2.50E-02 6.25E-02 As indicated in Table 4-2, the SQN inputs for valve failures (with exception of SRV-Water) are bounded by the generic analysis. Generic calculations that included water release SRVs are re-analyzed for the CT applicable to SQN. Note 1: PRA model of Record, rev. 1, CAFTA .rr file. SOV-FTC (9.54E-04) is from NUREG/CR-6928 as SQN does not model this valve/mode in the PRA. Note 2: AOVs are grouped into three categories, the most restrictive value is used for this analysis. Note 3: SRVs are split into two categories, steam release (6.76E-05) and water release (6.25E-02). Table 8-1 from the WCAP [Ref 5] does not differentiate between water and steam. Judging from the value used in the generic analysis the steam and water relief SRVs may have been treated together. 4.6.4 Valve Failure Probabilities, Beta and Gamma Factors Table 4 Generic and SQN Specific Beta, Gamma Factors Valve Type Parameter Generic Analysis (Valve fail-to-close, betaftc)SQN [Note 1] (Valve fail-to-close, betaftc) AOV betaftc-aov 0.1 1.63E-02 CKV betaftc-ckv 0.1 8.50E-03 MOV betaftc-mov 0.088 1.54E-02 SOV betaftc-sov 0.1 0.1 [Note 2] SRV-Steam betaftc-srvS 0.22 7.19E-02 SRV-Water betaftc-srvW 0.22 0.22 [Note 2] all valve types Due to Valve Transferring Open, betato 0.1 0.1 [Note 2] all valve types Due to Valve Transferring Open, gammato 0.5 0.5 [Note 2] Note 1 - The conservative generic valves for SOVs and SRV-Water failure-to-close is used in the SQN analysis. Note 2 - The conservative generic values for the beta and gamma values are used in the SQN analysis. The SQN parameters listed in Table 4-3 are bounded by those used in the generic analysis. 4.6.5 Spurious (Transfer) Open Probabilities and Beta Factors Table 4-4 Spurious Open and Beta Factor Values Parameter Component Description Value Source AOV XO Air-Operated Valve Probability AOV spuriously opens per hour 1.82E-07 NUREG/CR-5496 (Section 2.4) AOVBETA Beta factor - AOV spuriously opens 1.63E-02 Ptopre-aov Probability AOV spuriously opens during 4-hr CT 7.28E-07 Calculated Pto-aov Probability AOV spuriously opens during 168-hr CT 3.06E-05 Calculated CKV XO Check Valve Probability CKV spuriously opens per hour (Leakage) 2.96E-08 NUREG/CR-6928 (Table 5-1) Parameter Component Description Value Source CKVBETA Beta factor - CKV Fails to Remain Closed 3.0E-02 NUREG/CR-5496 (2.5.1.2) Ptopre-ckv Probability CKV spuriously opens during 4-hr CT 1.18E-07 Calculated Pto-ckv Probability CKV spuriously opens during 168-hr CT 4.97E-06 Calculated MAN XO Manual Valve Probability MAN spuriously opens per hour (Leak) 6.67E-08 NUREG/CR-6928 (Table 5-1) MANBETA Beta factor - MAN spuriously opens 0.1 WCAP (Generic) Ptopre-man Probability MAN spuriously opens during 4-hr CT 2.67E-07 Calculated Pto-man Probability MAN spuriously opens during 168-hr CT 1.12E-05 Calculated MOV XO Motor-Operated Valve Probability MOV spuriously opens per hour 4.45E-08 NUREG/CR-6928 (Table 5-1) MOVBETA Beta factor - MOV spuriously opens 2.67E-02 NUREG/CR-5496 (2.3.1.2) Ptopre-mov Probability MOV spuriously opens during 4-hr CT 1.78E-07 Calculated Pto-mov Probability MOV spuriously opens during 168-hr CT 7.48E-06 Calculated SRV XO Safety Relief Valve Probability SRV (Water or Steam Release) spuriously opens per hour 2.12E-07 Calculated SRVBETA-W Beta factor - SRV-Water fails to reseat after opening 0.22 WCAP SRVBETA-S Beta factor - SRV-Water fails to reseat after opening 0.22 WCAP Ptopre-srv Probability SRV-Water or Steam spuriously opens during 4-hr CT 8.48E-07 Section 4.6.1 Pto-srv Probability SRV-Water or Steam spuriously opens during 168-hr CT 3.56E-05 Section 4.6.1 SOV XO Solenoid Operated Valve Probability SOV spuriously opens per hour 9.23E-08 NUREG/CR-6928 (Table 5-1) Parameter Component Description Value Source SOVBETA Beta factor - SOV spuriously opens 0.1 WCAP (Generic) Ptopre-sov Probability SOV spuriously opens during 4-hr CT 3.69E-07 Calculated Pto-sov Probability SOV spuriously opens during 168-hr CT 1.55E-06 Calculated The values listed in Table 4-4 are bounded by those used in the generic analysis. 4.6.6 Additional Inputs Table 4-5 Additional Inputs Parameter Description Generic Value Value Source PBSEIS Seismic Pipe Break Probability for Non-Seismically Qualified Pipe 1.0 1.0 Assumed PBRANDOM Random Pipe Break Frequency (per year)1 1.10E-03 3.14E-03 Ref. 9 Table E-1 Ptopre Probability that Valve Spuriously Transfers Open During 4-Hr CT (most limiting valve) 4.00E-06 8.48E-07 Calculated Section 4.6.1 Pto Probability that Valve Spuriously Transfers Open During 168-Hr CT (most limiting valve) 1.68E-04 3.56E-05 Calculated Section 4.6.1 Pm1 Probability that a CIV is Disabled due to Maintenance (per demand) during a 4-hr CT 4.00E-06 2.28E-06 Calculated Section 4.6.1 PmE Probability that Extra Valve is Disabled due to Maintenance (per demand) [assume extra valve currently has 72 hour CT] Any Valve 4.11E-05 Calculated Section 4.6.1 Pm2 Probability that a CIV is Disabled due to Maintenance (per demand) during a 168-hr CT Any Valve 9.59E-05 Calculated Section 4.6.1 Note 1 - The random (passive) pipe break frequency is based on the most limiting frequency used in the SQN internal flooding analysis. The values listed in Table 4-5 are bounded by those used in the generic analysis with exception to the random pipe break frequency. Therefore, the generic analysis that include the random pipe break frequency are reanalyzed for SQN to determine the applicable CT.

4.7 Application Tiers 1, 2 and 3 4.7.1 Tier 1 PRA Applicability and Insights The SQN PRA was subjected to a full scope Peer Review in accordance with R.G. 1.200 [Ref 4] requirements in 2011. The conclusions of the peer review team follow: [Ref 13]

The overall model structure is robust and well-developed, but needs refinement,
Documentation is thorough, detailed and well-organized such that comparison with the standard is facilitated,
The process and tools utilized are at the state-of-technology and generally consistent with Capability Category II, and
The PRA maintenance and update program includes all necessary processes and does a very good job of tracking pending changes.

The following areas are addressed: 1. Assurance that the plant-specific PRA reflects the as-built, as-operated plant. A key attribute to the ASME/ANS Standard is to assess how the PRA modeled the as-built as-operated plant. The Data Analysis technical element addresses this. The PRA model at the time of the Peer Review was judged to meet this requirement, and HLR-MU-B stated that a PRA configuration control process is in place, and governed by procedure which provides a reasonable assurance that the as-built, as-operated planted is reflected through routine maintenance and upgrades to the PRA.

2. Assurance that the applicable PRA updates include the findings from the individual plant evaluation (IPE) and the IPE for External Events.

The SQN PRA has been updated multiple times since the completion of the IPE and IPEEE from the 1990's. The technical adequacy of the PRA was established by Peer Review in early 2011. The current model of record represents a significantly more mature PRA as compared to the IPE and IPEEE. 3. Assurance that conclusions from the peer review, including facts and observations that are applicable to this application have been resolved. For areas, whereby the Peer Review determined work was necessary to meet Capability Category II a Facts & Observation (F&O) was initiated. The resolutions to the F&Os are documented in the PRA Summary Notebook. [Ref 7]

4. Assurance that there is PRA configuration control and updating, including PRA quality assurance programs, associated procedures, and PRA revision schedules.

TVA procedure NPG-SPP-09.11, Probabilistic Risk Assessment (PRA) Program, covers the management of PRA applications and periodic PRA updates. Periodic changes made to the base plant-specific PRA model are required to incorporate system, structure, component and operating philosophy changes, and new plant-specific data.

5. Assurance that there is PRA adequacy, completeness, and applicability with respect to evaluating the risk associated with the proposed CIV CT extensions. SQN specific parameters and PRA results applicable to the proposed risk-informed application of CIV completion time extensions are well documented in references 7 through 10. The PRA has been subjected to a Peer Review in early 2011 that assessed the technical adequacy of the SQN PRA.

6. Assurance that plant design or operational modifications that are related to or could impact the proposed CT extensions are reflected in the PRA revision used in the plant-specific application, or a justification for not including those modifications in the PRA.

In accordance with TVA procedure NPG-SPP-09.11, "Probabilistic Risk Assessment (RPA) Program," plant modifications or design changes that result in new configurations, alignments, and capabilities of plant system are assessed for inclusion in model updates. Furthermore TVA procedure NEDP-26 "Probabilistic Risk Assessment (PRA)" provides the requirements for the cumulative impact of plant configuration changes, including plant-specific design, procedure and operational changes that require an update to the Model of Record. 4.7.2 Tier 2 Avoidance of Risk-Significant Plant Configurations The process SQN uses to avoid risk-significant plant configurations is governed by TVA procedure NPG-SPP-07.1, "On-Line Risk Management." The procedure applies to all work activities that affect or have the potential to affect a plant component, system, or unit configuration. A risk assessment methodology is used for on-line maintenance and shutdown operations. For on-line maintenance, a risk assessment is performed prior to implementation and emergent work is evaluated against the assessed scope. Shutdown risk is assessed in accordance with TVA procedure NPG-SPP-07.2, "Outage Management." Furthermore, TVA procedure NPG-SPP-07.3, "Work Activity Risk Management Process" provides an integrated process for assessing and reducing the likelihood and/or consequences of an adverse event. SQN employs a work management process that utilizes Functional Equipment Groups (FEGs). The grouping qualitatively assessed work activities and components and made logic ties that prevent certain risk-significant plant configurations for being scheduled simultaneously. 4.7.3 Tier 3 Risk-Informed Configuration Risk Management In accordance with the requirements of 10CFR50.65(a)(4) SQN assesses and manages plant configurations prior to taking the maintenance configuration. The proposed plant configuration is modeled in the computer code EOOS (Equipment Out Of Service) to determine the change in the core damage frequency (CDF) and the large early release frequency (LERF). The initial risk assessment is performed six - nine weeks prior to implementation to allow for risk-informed sequencing of activities as necessary and for other actions determined based on risk insights gleaned from the initial assessment. The well defined process is governed by TVA procedure NPG-SPP-07.1, "On-Line Risk Management." The quantified change in risk is used as one input with respect to configuration risk management. Furthermore, the process prescribes successive higher levels of management approval for plant configurations resulting in an increase in risk at various levels. Although not quantified, work management compensatory measures are prescribed as the risk level increases to limit the likelihood of entering an unplanned configuration (i.e., protected trains/equipment) or to limit the consequences of an unattended action. Outage Risk Management is controlled in accordance with TVA procedure NPG-SPP-7.2.11.

4.8 SQN Specific Analysis 4.8.1 Fault Trees and Applicable Penetrations The following fault trees were developing to calculate the given penetration configuration. Table 4-6 Class I Classification and Penetrations: SQN FT ID Class / Group Calculation Number Applicable Penetrations I_A-1_4 I, A#1 I,B#3 I,C#3 X-79A X-79B X-80 X-82 X-83 I_A-1_168 I_A-3_4 I,A#3 I,B#5 I, C#5 X-4 X-5 X-6 X-7 X-9A X-9B X-10A X-10B X-11 X-29 X-43A X-43B X-43C X-43D X-47A X-47B X-50A X-52 X-57 X-58 X-59 X-60 X-61 X-62 X-63 I_A-3_168 I_A-4-4 I,A#4 X-42 X-50B X-51 X-78 X-111 X-112 X-113 I_A-4-168 I_B-1-4 I,B#1 X-35 X-88 X-117 X-118 I_B-1-168 I_B-4-4 I,B#4 X-40D(U1) X-40D(U2) X-48A X-48B X-49A X-49B I_B-4-168 I_B-5-4 I,B#5 X-46 I_B-5-168 I_B-6-4 I,B#6 X-53 I_B-6-168 I_C-1-4 I,C#1 X-19A X-19B I_C-1-168

Table 4-7 Class II Classification and Penetrations: SQN FT ID Class / Group Calculation Number Applicable Penetrations II_A-6-4 II,A#6 X-20A X-20B II_A-6-168 II_A-17-4 II,A#17 X-17 X-21 X-32 II_A-17-168 II_B-2-4 II,B#2 X-15 II_B-2-168 II_B-X44-4 II,B#X44 X-44 II_B-X44-168 N/A II-Type A - Bounding X-22 X-33 X-107 N/A II-Type B - Bounding None Table 4-8 Class III Classification and Penetrations: SQN FT ID Class / Group Calculation Number Applicable Penetrations III_A-X12-4 III,A#X12 X-12B X-12C III_A-X12-8 III_A-X13-4 III,A#X13 X-13A X-13B X-13C X-13D III_A-X13-8 III_A-X14-4 III,A#14 X-14A X-14B X-14C X-14D III_A-X14-8 N/A III-Type A - Bounding X-12A X-12D N/A III-Type B - Bounding None Table 4-9 All Classes - Penetrations With CT Extensions Based on WCAP Generic Analyses Applicable Penetrations X-16 X-24 X-25C X-26C X-27D X-86A X-86B X-86C X-102 X-104 Table 4-10 Penetrations with One or More Valves Crediting the Small Line Exclusion Applicable Penetrations X-4 X-5 X-6 X-7 X-9A X-9B X-10A X-10B X-11 X-15 X-16 X-17 X-19A X-19B X-20A X-20B X-21 X-22 X-23 X-24 X-25A(U1) X-25A(U2) X-25B(U1) X-25B(U2) X-25D(U1) X-25D(U2) X-26A X-26B(U1) X-26B(U2) X-27A X-27B X-27C X-29 X-30 X-32 X-33 X-34(U1) X-34(U2) X-35 X-39A X-39B X-41 X-42 X-43A X-43B X-43C X-43D X-45 X-46 X-47A X-47B X-48A X-48B X-49A X-49B X-50A X-50B X-51 X-52 X-56 X-57 X-58 X-59 X-60 X-61 X-62 X-63 X-64 X-65 X-66 X-67 X-68(U2) X-69(U2) X-70(U2) X-71(U2) X-72(U2) X-73(U2) X-74(U2) X-75(U2) X-76(U1) X-76(U2) X-77 X-78 X-80 X-82 X-83 X-84A X-85B X-85C X-87B X-87D X-90(U1) X-90(U2) X-91 X-92A(U1) X-92A(U2) X-92B(U1) X-92B(U2) X-93(U1) X-93(U2) X-94A X-94B X-94C X-95A X-95B X-95C X-96C(U1) X-96C(U2) X-97 X-98 X-99(U1) X-99(U2) X-100(U1) X-100(U2) X-101 X-102 X-103 X-104 X-106 X-107 X-111 X-112 X-113 X-114 X-115 X-116A Table 4-11 Penetrations with No Generic Fit - Not Analyzed By PRA Applicable Penetrations X-40A X-40B X-108 X-109 4.8.2 Calculations / Inputs Two calculations are performed to determine the acceptability of the extended CT. The Incremental Conditional Large Early Release Probability (ICLERP) is based on R.G. 1.177 [Ref 3] acceptable criteria of <5.0E-08. The change in the large early release frequency (LERF) is based on the R.G. 1.174 [Ref 2] acceptance criteria of <1.0E-07/yr. a) Class I - penetrations connected to the containment atmosphere; a failure to isolate a penetration would result in a release path to the environment. Four types:

Type A - Flow paths connected directly to containment atmosphere and the outside environment.
Type B - Flow paths closed inside containment and connected directly to the outside environment.
Type C - Flow paths connected directly to containment atmosphere and closed outside containment.
Type D - Flow paths closed inside containment and closed outside containment.

Class I Penetrations - Flow Paths Connected to the Containment Atmosphere Release Type Details Comment Input(s) Non-seismic CDF Release For open systems a direct connection from inside to outside containment is possible given a failure to isolate the penetration. Release due to an internal event CDF. If core damage occurs simultaneous with a failure to isolate (spurious open, fail-to-close, etc.) the penetration a large release could occur.

CDFT
CT
Valve Failure Probability Seismic CDF Release For closed systems (either inside or outside of containment), a seismic-induced core damage event, the assumption is made that the closed loop system piping fails. Closed systems - seismic event breaches both sides of containment. If this were to occur simultaneous with failure to isolate the penetration, an open pathway to the environment would exist.
CDFseis
CT
PBSEIS
Valve Failure Probability Random Pipe Break CDF Release The configuration would be based on the system being open on one side of containment and the other closed. Release due to an internal event and a random pipe break. A random pipe break of the closed system simultaneous with a failure to isolate the penetration would present a flow path to the environment.
CDFT
CT
PBRANDOM
Valve Failure Probability b) Class II - penetration flow paths connected to the reactor coolant system. Two types:
Type A - Standby system flow paths.
Type B - Normally operating system flow paths. Class II Penetrations - Flow Paths Connected to the Reactor Coolant System Release Type Details Comment Input(s) Non-seismic CDF Release Connected to the RCS and open outside containment. Release due to an internal event CDF. Core damage simultaneous with CIV failure to isolate (spurious open, fail-to-close, etc.) the penetration a large release could occur.
CDFT
CT
Valve Failure Probability Seismic CDF Release For system connected to the RCS and open outside it is assumed that a seismic event results in core damage.

Release due to seismic event resulting in core damage whereby all closed loop piping fails both inside and outside containment simultaneous with CIV failure creating an opening to the environment.

CDFSEIS
CT
PBSEIS
Valve Failure Probability Release Type Details Comment Input(s) Random Pipe Break CDF Release For systems connected to the RCS and open outside containment. Release due to an internal event and a random pipe break. If core damage were to occur simultaneous with a random pipe break inside containment and a failure to isolate the penetration, the system would no longer be connected to the RCS, therefore, allowing an open flow path to the environment.
CDFT
CT
PBRANDOM
Valve Failure Probability Interfacing System Loss of Coolant Accident (ISLOCA) For standby systems connected to the RCS and open outside containment. Release due to containment bypass, if the CIVs fail (CIV failure is the initiator), an ISLOCA would occur resulting in core damage.
Valve Failure Probability c) Class III - penetrations with flow paths connected to the steam generators. Two types:
Type A - Flow paths connected to the steam generator secondary side and open to the outside environment.
Type B - Flow paths connected to the steam generator secondary side and closed to the outside environment. Class III Penetrations - Flow Paths Connected to the Steam Generator Secondary Side Release Type Details Comment Input(s) Seismic CDF Release For systems connected to the SG secondary side and open or closed outside containment.

Release due to seismic event resulting in core damage whereby all closed loop piping fails both inside and outside containment simultaneous with CIV failure creating an opening to the environment.

CDFSEIS
CT
PBSEIS
Valve Failure Probability Random Pipe Break CDF Release For systems connected to the SG secondary side and open outside containment. Release due to an internal event and a random pipe break. If core damage were to occur simultaneous with a random pipe break inside containment and a failure to isolate the penetration, the system would no longer be connected to the SGs, therefore, allowing an open flow path to the environment.
CDFT
CT
PBRANDOM
Valve Failure Probability Release Type Details Comment Input(s) Steam Generator Tube Rupture (SGTR) For systems connected to the steam generator secondary side and open to the outside atmosphere. Release due to SGTR simultaneous with a core damage event and failure to isolate the penetration which would result in an open pathway to the environment
CDFSGTR
Valve Failure Probability
CT Steam Generator Tube Rupture (SGTR)

With Random Pipe Break For systems connected to the SG secondary side and a closed systems outside containment. Random pipe break outside of containment followed by a SGTR and CIV failure would result in an open path to the environment.

CDFSGTR
PBRANDOM
Valve Failure Probability
CT 5.0 Results The results of the generic analysis and the SQN specific analysis are recorded in Attachment 1. Many CIVs were justified at 168-hr CTs based on application of the generic analysis. Justification could not be made for some CIVs, therefore their CTs remain at 4-hrs.

ATTACHMENT 1 Page 1 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-4 30-56 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-57 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-555TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-5 30-58 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-59 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-554TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-6 30-50 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 2 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 30-51 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-6 (cont) 30-558TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-7 30-52 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-53 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-557TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-9A 30-8 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-7 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-563TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs ATTACHMENT 1 Page 3 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-9B 30-10 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-9B (cont) 30-9 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-562TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-10A 30-15 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-14 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-561TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-10B 30-17 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 4 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 30-16 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-560TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-11 30-20 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-19 This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Assume more limiting condition of valve open. Same valve type IC and OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-559TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Normally closed valve. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-12A 3-33 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.3, MFIV. N/A 3-164 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A 3-164A THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A 3-174 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A ATTACHMENT 1 Page 5 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 3-904 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-903 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-857 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-12A (cont) 3-889 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-849 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-853 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs U2 ONLY 2-3-504 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-12B 3-47 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.3, MFIV. N/A 3-502 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. Drain valve assume same CT as 3-47. III-A-12BC System pressure boundary maintained System pressure boundary compromised 8-hrs 8-hrs 8-hrs 8-hrs 8-hrs 8-hrs X-12C 3-87 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.3, MFIV. N/A ATTACHMENT 1 Page 6 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 3-500 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. Drain valve assume same CT as 3-87. III-A-12BC System pressure boundary maintained System pressure boundary compromised 8-hrs 8-hrs 8-hrs 8-hrs 8-hrs 8-hrs X-12D 3-100 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.3, MFIV. N/A 3-171 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A 3-171A THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A X-12D (cont) 3-175 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A 3-907 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-906 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-858 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-890 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-850 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 3-854 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs ATTACHMENT 1 Page 7 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT UNIT 2 ONLY 2-3-506 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III,A Bounding System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-13A 1-5 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.4, ARV. N/A 1-4 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.2, MSIV. N/A 1-147 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-13A (cont) 1-15 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.5, AFW. N/A 1-522 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-523 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-524 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-525 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-526 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-922 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs ATTACHMENT 1 Page 8 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 1-536 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-13B 1-12 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.4, ARV. N/A 1-11 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.2, MSIV. N/A 1-148 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-13B (cont) 1-517 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-518 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-519 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-520 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-521 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-923 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-534 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs ATTACHMENT 1 Page 9 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-13C 1-23 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.4, ARV. N/A 1-22 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.2, MSIV. N/A 1-149 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-512 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A X-13C (cont) 1-513 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-514 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-515 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-516 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-924 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-532 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-13D 1-30 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.4, ARV. N/A ATTACHMENT 1 Page 10 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 1-16 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.3, MFIV. N/A 1-150 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-29 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.2, MSIV. N/A 1-527 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A X-13D (cont) 1-528 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-529 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-530 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-531 THIS VALVE IS NOT COVERED BY ITS SECTION 3.6.3. IT IS COVERED BY ITS SECTION 3.7.1, MSSV. N/A 1-925 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-538 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-13 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-14A 1-182 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve IC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs ATTACHMENT 1 Page 11 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 1-14 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 43-58 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-825 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-14B 1-184 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve IC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-14B (cont) 1-32 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 43-64 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-827 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-14C 1-183 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve IC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-25 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 43-61 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-826 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs ATTACHMENT 1 Page 12 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-14D 1-181 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve IC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-7 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 43-55 Direct connection to Steam Generator. Closed system IC to open system OC. Normally open valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 1-824 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed valve OC. III-A-14 System pressure boundary maintained System pressure boundary compromised 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs 8-Hrs X-15 62-72 Normally operating system; RCS connection. 3 Valves IC, 1 normally open, 2 normally closed. 1 Valves OC, normally open. All Valves the same type. This valve normally closed IC. II-B-2 System pressure boundary maintained System pressure boundary compromised 168-Hrs168-Hrs 168-Hrs168-Hrs168-Hrs168-Hrs 62-73 Normally operating system; RCS connection. 3 Valves IC, 1 normally open, 2 normally closed. 1 Valves OC, normally open. All Valves the same type. This valve normally open IC. II-B-2 System pressure boundary maintained System pressure boundary compromised 168-Hrs168-Hrs 168-Hrs168-Hrs168-Hrs168-Hrs 62-74 Normally operating system; RCS connection. 3 Valves IC, 1 normally open, 2 normally closed. 1 Valves OC, normally open. All Valves the same type. This valve normally closed IC. II-B-2 System pressure boundary maintained System pressure boundary compromised 168-Hrs168-Hrs 168-Hrs168-Hrs168-Hrs168-Hrs ATTACHMENT 1 Page 13 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-662 Normally operating system; This valve is a PRESSURE RELIEF VALVE, WHICH RELIEVES TO THE PRESSURIZER RELIEF TANK INSIDE CONTAINMENT and is not directly connected to the RCS, Given this scenario, flow path is also smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-77 Normally operating system; RCS connection. 3 Valves IC, 1 normally open, 2 normally closed. 1 Valves OC, normally open. All Valves the same type. This valve normally open OC. II-B-2 System pressure boundary maintained System pressure boundary compromised 168-Hrs 168-Hrs 168-Hrs168-Hrs 168-Hrs 168-Hrs 62-707 This is a normally closed test valve which communicates with containment atmosphere not RCS for flow thru valve from IC to OC. Therefore flow path is smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-16 62-543 Normally operating system; RCS connection. 1 CIV IC normally open - 1 CIV OC, normally open - different valve types (The normally open check valve IC has another normally open check valve in series between it and the RCS) (The normally open CIV OC has another normally open valve downstream of it, same valve type) II,B #3 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-90 Normally operating system; RCS connection. 1 CIV IC normally open - 1 CIV OC, normally open - different valve types (The normally open check valve IC has another normally open check valve in series between it and the RCS) (The normally open CIV OC has another normally open valve downstream of it, same valve type) II,B #3 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 14 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-709 Normally operating system; RCS connection. Continues to operate during accident, therefore not considered a path for release directly from RCS since flow continues to be forced into RCS; therefore, release scenario is from containment atmosphere, flow path is smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-544 This is a normally closed test valve which communicates with containment atmosphere not RCS for flow thru valve from IC to OC. Therefore flow path is smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-17 63-640 RCS connection; standby system. Open system IC. Closed system OC. Normally closed valve IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-643 RCS connection; standby system. Open system IC. Closed system OC. Normally closed valve IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs X-17 (cont) 63-158 No direct connection to RCS piping; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-172 RCS connection; standby system. Open system IC. Closed system OC. Normally closed valve IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-637 No direct connection to RCS piping; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-636 No direct connection to RCS; Only release path is from containment atmosphere to environment via RHA. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 15 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-642 No direct connection to RCS; Only release path is from containment atmosphere to environment via RHA. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-870 No direct connection to RCS; Only release path is from containment atmosphere to environment via RHA. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-19A 63-72 Standby system. Containment atmosphere at sump. Closed system OC. 1 valve - normally closed (OC or IC) I-C-1 System pressure boundary maintained Pressure Boundary Compromised 168-hrs 4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-593 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-591 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-19B 63-73 Standby system. Containment atmosphere at sump. Closed system OC. 1 valve - normally closed (OC or IC) I-C-1 System pressure boundary maintained Pressure Boundary Compromised 168-hrs 4-hrs 168-hrs168-hrs 168-hrs 4-hrs 63-592 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-590 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-20A 63-112 No direction connection to RCS piping; Line isolated by 2 normally closed valves. Valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs ATTACHMENT 1 Page 16 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-635 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 63-633 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 63-94 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 63-631 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-20A (cont) 63-667 No direction connection to RCS piping; valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-661 No direction connection to RCS piping; valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 17 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-833 No direction connection to RCS piping; valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-20B 63-111 No direction connection to RCS piping; Line isolated by 2 normally closed valves. Valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-632 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs168-hrs 168-hrs168-hrs 168-hrs 168-hrs 63-634 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-20B (cont) 63-93 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs168-hrs 168-hrs168-hrs 168-hrs 168-hrs 63-630 RCS connection; standby system. 2 check valves IC each have another normally closed check valve in series. 2 valves OC in parallel. 1 normally open and 1 normally closed. II-A-6 System pressure boundary maintained System pressure boundary compromised 168-hrs168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 18 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-413 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 63-659 No direction connection to RCS piping; valve is IC; only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 63-660 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-21 63-167 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 63-547 RCS connection; standby system; normally closed check valve IC; another check valve upstream IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs 4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-549 RCS connection; standby system; normally closed check valve IC; another check valve upstream IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs X-21 (cont) 63-157 No Direct connection to RCS; Open system IC to closed system OC; Normally closed valve OC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-648 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 19 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-649 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-650 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-862 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-313A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-314A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs X-21 (cont) 63-317A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 63-318A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs ATTACHMENT 1 Page 20 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-22 63-174 Connected to RCS accumulators thru another normally closed FCV. Not directly connected to RCS. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 168 hrs 63-581 RCS connection; standby system. Open system IC and closed system OC. Normally closed valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-25 (FCV) RCS connection; standby system. Open system IC and closed system OC. Normally closed valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-26 (FCV) RCS connection; standby system. Open system IC and closed system OC. Normally closed valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs Unit 2 Only 2-63-816 Not directly connected to RCS. Drain or vent valve. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-25 (FSV) Not directly connected to RCS. Open system IC to closed system OC. Normally closed valve OC. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-26 (FSV) Not directly connected to RCS. Open system IC to closed system OC. Normally closed valve OC. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-23 43-310 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-309 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 21 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 43-492 Test valve connected to containment atmosphere IC. Only release path is from containment atmosphere to environment via the sampling system. Source piping is 3/8" and smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-24 68-559 No direct connection to RCS; penetration flow path connects open system IC to closed system OC. Normally closed valve IC & OC. Different valve types. This valve IC. I,A#6 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-505 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-512 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-513 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-511 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-24 (cont) 63-536 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-535 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-534 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 22 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-626 No direct connection to RCS; penetration flow path connects open system IC to closed system OC. Normally closed valve IC & OC. Different valve types. This valve OC. I,A#6 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-627 No direct connection to RCS; penetration flow path connects open system IC to closed system OC. Normally closed valve IC & OC. Different valve types. This valve OC. I,A#6 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 68-560 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 68-561 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-517 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-518 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-638 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-25A 43-2 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 23 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 43-3 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-25B - THE DP SENSORS ARE CLOSED SYSTEMS OUTSIDE OF CONTAINMENT THAT ARE ATTACHED DIRECTLY TO CONTAINMENT. NO ISOLATION VALVES ARE EMPLOYED FOR THESE SENSORS AS THEY USE A DOUBLE DIAPHRAGM SYSTEM FOR DP MEASUREMENT. THE DIAPHRAGMS ARE QUALIFIED FOR POST-LOCA USE. NO DIRECT CONNECTION TO RCS. FLOW PATH SMALLEER THAN THAT REQUIRED TO RESULT IN A LARGE RELEASE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-311Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-311X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-44Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-25B (cont) 30-44X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 24 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-25C - THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-25D 43-11 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-12 Lines connected to the RCS 3/8" in diameter or less are within the makeup capability of plant's charging systems and therefore, are not considered small LOCAs or potential containment bypass pathways. (Sec 8.2.3 of WCAP) Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 25 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-26A - THE DP SENSORS ARE CLOSED SYSTEMS OUTSIDE OF CONTAINMENT THAT ARE ATTACHED DIRECTLY TO CONTAINMENT. NO ISOLATION VALVES ARE EMPLOYED FOR THESE SENSORS AS THEY USE A DOUBLE DIAPHRAGM SYSTEM FOR DP MEASUREMENT. THE DIAPHRAGMS ARE QUALIFIED FOR POST-LOCA USE. NO DIRECT CONNECTION TO RCS. FLOW PATH SMALLEER THAN THAT REQUIRED TO RESULT IN A LARGE RELEASE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-310Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-310X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-43Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-43X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-26B Unit 1 Only 32-102 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-297 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-295 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 26 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-26B Unit 1 Only (cont) 32-292 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-26B Unit 2 Only 32-103 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-348 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-341 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-345 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-26C - THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 27 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-27A - THE DP SENSORS ARE CLOSED SYSTEMS OUTSIDE OF CONTAINMENT THAT ARE ATTACHED DIRECTLY TO CONTAINMENT. NO ISOLATION VALVES ARE EMPLOYED FOR THESE SENSORS AS THEY USE A DOUBLE DIAPHRAGM SYSTEM FOR DP MEASUREMENT. THE DIAPHRAGMS ARE QUALIFIED FOR POST-LOCA USE. NO DIRECT CONNECTION TO RCS. FLOW PATH SMALLEER THAN THAT REQUIRED TO RESULT IN A LARGE RELEASE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-30CX No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-30CY No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-27B - THE DP SENSORS ARE CLOSED SYSTEMS OUTSIDE OF CONTAINMENT THAT ARE ATTACHED DIRECTLY TO CONTAINMENT. NO ISOLATION VALVES ARE EMPLOYED FOR THESE SENSORS AS THEY USE A DOUBLE DIAPHRAGM SYSTEM FOR DP MEASUREMENT. THE DIAPHRAGMS ARE QUALIFIED FOR POST-LOCA USE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-42Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-42X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 28 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-27C 52-504 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-27C (cont) 52-505 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 52-510 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-27D - THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-29 70-89 No direct connection to RCS; penetration flow path connects open system IC to open system OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-698 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-92 No direct connection to RCS; penetration flow path connects open system IC to open system OC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 29 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 70-735 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-30 63-71 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-84 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-23 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-537 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-344A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-32 63-21 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-545 RCS connection; standby system; normally closed check valve IC; another check valve upstream IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-543 RCS connection; standby system; normally closed check valve IC; another check valve upstream IC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs 63-156 No Direct connection to RCS; Open system IC to closed system OC; Normally closed valve OC. II-A-X17 System pressure boundary maintained System pressure boundary compromised 168-hrs4-hrs 168-hrs4-hrs 168-hrs 4-hrs ATTACHMENT 1 Page 30 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-541 No direct connection to RCS; isolated from RCS by double check valves. Open system IC to closed system OC. Normally closed valve OC. Drain or vent valve. Assume same CT as the shortest CT of other valves in the penetration. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-32 (cont) 63-823 No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-657 No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-658 No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-864 No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-315A No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 31 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-316A No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-311A No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-32 (cont) 63-612A No direct connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-33 63-551 RCS connection; standby system; open system IC to closed system OC. Normally closed valve IC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-553 RCS connection; standby system; open system IC to closed system OC. Normally closed valve IC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-555 RCS connection; standby system; open system IC to closed system OC. Normally closed valve IC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-557 RCS connection; standby system; open system IC to closed system OC. Normally closed valve IC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-121 No direction connection to RCS, standby system. Isolated from RCS by at least 2 normally closed valves. Open system IC to closed system OC. Normally closed valve IC. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 32 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-22 Penetration flow path connects OPEN system IC to closed system OC. Normally open valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 63-653 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-33 (cont) 63-654 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-655 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-836 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-656 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-831 No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 33 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-325A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-326A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-33 (cont) 63-319A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-320A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-321A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-322A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-323A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 34 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 63-324A No direction connection to RCS piping; Valve is IC; Only release path is from containment atmosphere to environment via the SIS system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-34 Unit 1 Only 32-377 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-110 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-34 Unit 1 Only (cont) 32-375 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-373 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-34 Unit 2 Only 32-387 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-111 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-385 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-383 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-35 70-85 No direct connection to RCS; penetration flow path connects closed system IC to open system OC. I-B-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-703 No direct connection to RCS; penetration flow path connects closed system IC to open system OC. I-B-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 35 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 70-702C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-762 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-702F No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-764 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-35 (cont) Unit 2 Only 2-70-759 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-39A 77-868 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 63-64 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-867 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-39B 77-849 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 68-305 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-848 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 36 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-40A 3-156 THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-156A THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-173 THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-860 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs X-40A (cont) 3-899 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-900 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-852 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-848 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-888 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-901 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs ATTACHMENT 1 Page 37 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-40B 3-148 THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-148A THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-172 THIS VALVE IS NOT COVERED BY TECH SPEC 3.6.3. IT IS COVERED BY TECH SPEC SECTION 3.7.5, AFW. N/A 3-859 Flow thru this valves is from containment atmosphere IC to OC. Flow path smaller in size than that required to result in a large release. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs X-40B (cont) 3-842 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-897 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-896 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-855 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-847 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 3-851 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs ATTACHMENT 1 Page 38 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 3-887 Direct connection to Steam Generator. Closed system IC to open system OC. Normally closed drain/vent valve OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs 4 hrs X-40D BLF No direct connection to RCS; penetration flow path connects open system IC to open system OC. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-41 77-127 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-128 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-42 81-502 No direct connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve IC & OC. Different valve type. This valve IC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-42 (cont) 81-12 No direct connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve OC & IC. Different valve type. This valve OC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 81-529 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43A 62-563 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-550 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 39 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-549 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-546 No direct connection to RCS; flow path smaller in size than that required to result in a large release. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-578 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-555 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43A (cont) 62-571 No direct connection to RCS; Connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-575 No direct connection to RCS; Connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43B 62-561 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 40 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-550 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-549 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43B (cont) 62-546 No direct connection to RCS; Normally operating system, continues to operate during accident. Therefore, not considered a path for release directly for RCS since flow continues to be forced into RCS; therefore release scenario is from containment atmosphere. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-577 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 41 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-555 No direct connection to RCS; Normally closed vent/drain valve. Isolated from RCS by double check valves. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-569 No direct connection to RCS; Vent/Drain connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-573 No direct connection to RCS; Vent/Drain connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43C 62-562 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-43C (cont) 62-550 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 42 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-549 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-546 No direct connection to RCS; flow path smaller in size than that required to result in a large release. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-579 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-555 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43C (cont) 62-570 No direct connection to RCS; Connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-574 No direct connection to RCS; Connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 43 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-43D 62-560 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-550 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-549 No direct connection to RCS; Normally operating system, continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore, release scenario is from containment atmosphere. Flow path is smaller in size than that required to result in a large release. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-43D (cont) 62-546 No direct connection to RCS; Normally operating system, continues to operate during accident. Therefore, not considered a path for release directly for RCS since flow continues to be forced into RCS; therefore release scenario is from containment atmosphere. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 44 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 62-576 Direct connection to the RCS. Normally operating system and continues to operate during accident, therefore not considered a path for release directly for RCS since flow continues to be forced into RCS. Therefore release scenario is from containment atmosphere. Open system IC to Closed system OC. Normally open valve IC. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-555 No direct connection to RCS; Normally closed vent/drain valve. Isolated from RCS by double check valves. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-568 No direct connection to RCS; Vent/Drain connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 62-572 No direct connection to RCS; Vent/Drain connected to containment atmosphere. Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-44 62-61 Direct connection to RCS; Normally operating system, penetration flow path connects open system IC to open system OC; normally open valve IC. II-B-X44 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 62-639 Direct connection to RCS; Normally operating system, penetration flow path connects open system IC to open system OC; normally closed check valve IC. II-B-X44 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-44 (cont) 62-63 Direct connection to RCS; Normally operating system, penetration flow path connects open system IC to open system OC; normally open valve OC. II-B-X44 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 45 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-45 77-18 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-19 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-20 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 77-984 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-46 77-9 Normally operating system; RCS connection, however b/c of relief valve on RC drain tank and pump discharge pressure IC (SQN Dwg 47E8330-1), extremely unlikely to reach RCS pressure, therefore considered as containment atmosphere connection from open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-B-5 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 77-10 Normally operating system; RCS connection, however b/c of relief valve on RC drain tank and pump discharge pressure IC (SQN Dwg 47E8330-1), extremely unlikely to reach RCS pressure, therefore considered as containment atmosphere connection from open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an OC valve. I-B-5 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 46 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-46 (cont) 84-511 Normally operating system; RCS connection, however b/c of relief valve on RC drain tank and pump discharge pressure IC (SQN Dwg 47E8330-1), extremely unlikely to reach RCS pressure, therefore given this scenario, flow path is also smaller in size than that required to result in a large release Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-47A 61-191 No connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 61-192 No connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 61-533 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-532 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-47B 61-193 No connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 61-194 No connection to RCS; penetration flow path connects open system IC to open system OC; normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 61-680 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 47 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 61-681 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-48A 72-547 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an IC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-39 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Same valve type. This is an OC valve. . I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-545 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-543 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-48B 72-548 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC. & OC. Different valve type. This is an IC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-2 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an OC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-546 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-544 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-49A 72-556 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an IC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 48 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 72-40 (FCV) No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an OC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-49A (cont) 72-215E No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-216E No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-215F No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-216F No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-40 (RFV) No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-552 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-49B 72-555 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an IC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-41 (FCV) No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC & OC. Different valve type. This is an OC valve. I-B-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 72-217E No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 49 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 72-218E No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-217F No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-49B (cont) 72-218F No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-41 (RFV) No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 72-551 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-50A 70-87 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-687 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-90 No direct connection to RCS; penetration flow path connects open system IC to open system. OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-737 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-50B 70-679 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Different valve type. This is an IC valve. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 50 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 70-134 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Different valve type. This is an OC valve. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-678B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-51 26-1260 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally closed valve IC. NO valve OC. Different valve types. This is an IC valve. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 26-240 No direct connection to RCS; penetration flow path connects open system IC to closed system OC; normally open valve OC. Normally closed IC. Different valve types. This is an OC valve. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 26-1258 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-52 70-791 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-140 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-141 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-691B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 51 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-53 70-143 No direct connection to RCS; penetration flow path connects closed system IC to open system OC. I-B-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-703 No direct connection to RCS; penetration flow path connects closed system IC to open system OC. I-B-6 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 70-760 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-53 (cont) 70-702B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-765 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 70-702E No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-54 BLF No direct connection to RCS; penetration flow path connects open system IC to open system OC. Assume blind flanges to be normally closed valves. Same type. I-A-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 72-hrs 168-hrs168-hrs 168-hrs 72-hrs X-56 67-1523D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-83 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-89 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve.. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 52 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 67-772 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-561D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-57 67-575D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-57 (cont) 67-111 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-112 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve.. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-58 67-1523A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-107 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-106 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-778 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-561A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 53 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-59 67-575A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-87 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. . I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-88 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. . I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-60 67-1523B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-90 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-91 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-774 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-561B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-61 67-575B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-103 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 54 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 67-104 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-62 67-1523C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-99 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-62 (cont) 67-105 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-776 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-561C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-63 67-575C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-95 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve IC & OC. Same valve type. This is an IC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 67-96 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC & IC. Same valve type. This is an OC valve. I-A-3 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs X-64 31C-752 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 55 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 31C-223 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-222 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-65 31C-734 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-225 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-65 (cont) 31C-224 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-66 31C-715 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-230 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-229 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-67 31C-697 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-232 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 31C-231 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 56 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-68 Unit 2 Only 67-580D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-141 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-578D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-69 Unit 2 Only 67-580A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-130 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-69 Unit 2 Only (cont) 67-579A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-70 Unit 2 Only 67-585B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-297 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-139 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-71 Unit 2 Only 67-585C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-296 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 57 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 67-134 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-72 Unit 2 Only 67-585D No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-298 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-142 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-73 Unit 2 Only 67-585A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-73 Unit 2 Only (Cont) 67-295 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-131 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-74 Unit 2 Only 67-580B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-138 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-579B No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-75 Unit 2 Only 67-580C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 58 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 67-133 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 67-579C No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-76 Unit 1 Only 33-704 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 33-740 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 33-212 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-76 Unit 2 Only (cont) 33-722 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 33-739 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 33-211 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-77 59-633 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 59-522 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 59-529 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 59 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 59-704 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 59-651 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-78 26-1296 No direct connection to RCS; penetration flow path connects open system IC to open system OC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 26-243 No direct connection to RCS; penetration flow path connects open system IC to open system OC. Normally open valve OC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 26-1293 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-79A BLF No direct connection to RCS; penetration flow path connects open system IC to open system OC. 1 flange IC and 1 flange OC used to isolate the penetration and analyze as normally closed valves of the same type. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs X-79B BLF No direct connection to RCS; penetration flow path connects open system IC to open system OC. 1 flange IC and 1 flange OC used to isolate the penetration and analyze as normally closed valves of the same type. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs X-80 30-40 No direct connection to RCS. This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Open system IC to open system OC. Normally closed valve IC & OC. Same valve type. This is an IC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 60 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 30-37 No direct connection to RCS. This valve is normally closed but is intermittently opened to provide for containment min-purge during power operation. Open system IC to open system OC. Normally closed valve OC & IC. Same valve type. This is an OC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-556TP No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-82 78-560 No direct connection to RCS; Penetration flow path connects open system IC to open system OC. Normally closed valve IC & OC. Same valve type. This is an IC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs 78-561 No direct connection to RCS; Penetration flow path connects open system IC to open system OC. Normally closed valve OC & IC. Same valve type. This is an OC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs 78-228A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-83 78-558 No direct connection to RCS; Penetration flow path connects open system IC to open system OC. Normally closed valve IC & OC. Same valve type. This is an IC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs 78-557 No direct connection to RCS; Penetration flow path connects open system IC to open system OC. Normally closed valve OC & IC. Same valve type. This is an OC valve. I-A-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs 78-226A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-84A 68-308 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 61 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 68-307 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-85B THE DP SENSORS ARE CLOSED SYSTEMS OUTSIDE OF CONTAINMENT THAT ARE ATTACHED DIRECTLY TO CONTAINMENT. NO ISOLATION VALVES ARE EMPLOYED FOR THESE SENSORS AS THEY USE A DOUBLE DIAPHRAGHM SYSTEM FOR DP MEASUREMENT. THE DIAPHGRAMS ARE QUALIFIED FOR POST-LOCA USE. NO DIRECT CONNECTION TO RCS. FLOW PATH SMALLER IN SIZE THAN THAT REQUIRED TO RESULT IN A LARGE RELEASE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-45Y No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-45X No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 62 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-85C - THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. FLOW PATH SMALLER IN SIZE THAN THAT REQUIRED TO RESULT IN A LARGE RELEASE. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-86A - THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 63 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-86B - THIS LINE TRANSMITS PRESSURE FROM THE THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-86C - THIS LINE TRANSMITS PRESSURE FROM THE THIS LINE TRANSMITS PRESSURE FROM THE PRIMARY SYSTEM TO PRESSURE INSTRUMENTATION. THE LINE IS FLUID FILLED AND DOUBLE DIAPHRAGMED TO PREVENT COMMUNICATION BETWEEN THE PRIMARY SYSTEM FLUID AND THE AUXILIARY BUILDING. NO PRIMARY SYSTEM FLUID TRAVELS THROUGH THE PENETRATION SINCE THE INNER DIAPHRAGM IS LOCATED NEAR THE REACTOR VESSEL. SINCE DOUBLE DIAPHRAGMS ARE EMPLOYED FOR CONTAINMENT ISOLATION, NO CONTAINMENT ISOLATION VALVES ARE USED. II,A#9 System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 64 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-87B 52-502 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 52-503 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-87D 52-500 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 52-501 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-88 BLF No direct connection to RCS; penetration flow path connects open system IC to open system OC. Assume normally closed valve IC and OC. Same valve type. I-B-1 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs X-90 Unit 1 Only 32-287 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-80 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-285 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-281 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-90 Unit 2 Only 32-358 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-81 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 65 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-90 Unit 2 Only (cont) 32-353 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 32-354 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-91 43-251 Direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-250 Direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-497 Direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-92A, X-92B Unit 1 Only 43-207 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-452 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-424 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-208 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-453 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-423 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 66 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-92A, X-92B Unit 2 Only 43-207 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-210A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-525 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-417 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-208 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-210I No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-424 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-421 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-93 43-34 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-35 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-94A 90-109 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 67 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 90-107 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-94B 90-108 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 90-107 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-94C 90-110 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 90-111 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-95A 90-115 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 90-113 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-95B 90-114 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 90-113 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-95C 90-116 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 90-117 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 68 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-96C 43-22 Direct connection to RCS; Flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-23 Direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-97 30-134 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-135 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-98 52-506 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 52-507 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 52-508 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-99, X-100 Unit 1 Only 43-202 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-451 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-425 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-201 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 69 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 43-450 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-426 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-99, X-100 Unit 2 Only 43-202 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-200I No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-426 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-423 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-201 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-200A No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-427 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-419 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-101 43-319 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 70 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 43-318 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-474 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-102 3-352C This line joins to the secondary side of the steam generator inside containment and is considered a closed system inside containment. Direct connection Closed system IC to open system OC. Normally closed valve OC. III,A #1 System pressure boundary maintained System pressure boundary compromised 8 hrs 8 hrs 72 hrs 72 hrs 8 hrs 8 hrs Unit 2 Only 2-3-972 This valve is normally isolated from SG by valves 352A and 352B. Therefore flow is from containment atmosphere inside IC to OC. Open system IC to open system OC. Normally closed valve IC. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-103 43-461 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-317 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-341 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-464 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-104 3-351C This line joins to the secondary side of the steam generator inside containment and is considered a closed system inside containment. Direct connection Closed system IC to open system OC. Normally closed valve OC. III,A #1 System pressure boundary maintained System pressure boundary compromised 8 hrs 8 hrs 72 hrs 72 hrs 8 hrs 8 hrs ATTACHMENT 1 Page 71 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT Unit 2 Only 2-3-970 This valve is normally isolated from SG by valves 352A and 352B. Therefore flow is from containment atmosphere inside IC to OC. Open system IC to open system OC. Normally closed valve IC. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-104 (cont) Unit 2 Only 2-3-971 This valve is normally isolated from SG by valves 352A and 352B. Therefore flow is from containment atmosphere inside IC to OC. Open system IC to open system OC. Normally closed valve IC. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-106 43-460 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-325 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-307 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-469 No direct connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-107 74-2 RCS connection; standby system. Open system IC and OC. Normally closed valve IC downstream of another normally closed IC valve. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 74-1 RCS connection; standby system. Normally closed valve IC downstream of another normally closed IC valve. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 74-505 Standby system, no direct RCS connection. Relief valve discharges to the pressurizer relief tank which does not reach RCS pressure. Therefore, flow path is smaller than minimum size required for a large release. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs ATTACHMENT 1 Page 72 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 74-504 No direct connection to RCS piping. Valve is IC. Only release path is from containment atmosphere to environment via the RHR system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-107 (cont) 74-503 No direct connection to RCS piping. Valve is IC. Only release path is from containment atmosphere to environment via the RHR system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 74-549 No direct connection to RCS piping. Valve is IC. Only release path is from containment atmosphere to environment via the RHR system. Flow path is smaller than minimum size required for a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-108 BLF No connection to RCS; penetration flow path connects open system IC to open system OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs X-109 BLF No connection to RCS; penetration flow path connects open system IC to open system OC. Not Analyzed System pressure boundary maintained System pressure boundary compromised 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs 4-Hrs X-111 30-46 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 Pressure Boundary Maintained Pressure Boundary Compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-571 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 Pressure Boundary Maintained Pressure Boundary Compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs ATTACHMENT 1 Page 73 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT 30-46AX No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-46AY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-111 (cont) 30-46BY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-112 30-47 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 Pressure Boundary Maintained Pressure Boundary Compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-572 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 Pressure Boundary Maintained Pressure Boundary Compromised 168-hrs 168-hrs 168-hrs168-hrs 168-hrs 168-hrs 30-47AX No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-47AY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-47BY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 74 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-113 30-48 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 30-573 No RCS connection. The containment vacuum relief isolation butterfly valve is located in series with the vacuum relief valve (spring loaded check valve) all outside of the containment. Open system IC and OC. Normally open valve OC. I-A-4 System pressure boundary maintained System pressure boundary compromised 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs 168-hrs X-113 (cont) 30-48AX No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-48AY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 30-48BY No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-114 61-122 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-745 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-110 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-746 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs ATTACHMENT 1 Page 75 of 75 SQN Containment Isolation Valve Completion Time Results SQN Pent # SQN Valve Grouping Explanation Group & Calc # (Note 1) Maintenance Activity Type ICLERP @ CT: LERF @ CT: Justified CT X-115 61-97 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-692 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-96 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 61-691 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-116A 43-288 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-116A (cont) 43-287 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs 43-477 No connection to RCS; flow path smaller in size than that required to result in a large release. Small Line System pressure boundary maintained System pressure boundary compromised 168 hrs 168 hrs 168 hrs 168 hrs168 hrs 168 hrs X-117 BLF No connection to RCS; penetration flow path connects open system IC to open system OC. I-B-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs X-118 BLF No connection to RCS; penetration flow path connects open system IC to open system OC. I-B-1 System pressure boundary maintained System pressure boundary compromised 72-hrs 72-hrs 168-hrs168-hrs 72-hrs 72-hrs Note 1 - Group/Calc # such as I,A#6, II,A#9, II,B#3, etc. match the generic configurations and use the CT times from the generic calculations for WCAP-15791-P-A, Revision 2. Group/Clac# such as I-A-1, I-B-3, II-B-2, etc. match the generic configurations of the WCAP but the CT times were determined by calculations using SQN specific PRA values. Group/Calc # III-A-12BC, III-A-13, III-A-14, II-A-X17, II-B-X44, II-A-BOUNDING do not match generic configurations of the WCAP and have been analyzed and CT times determined using SQN specific configurations and SQN specific PRA values. The evaluations for the CTs for all the CIVs are documented in TVAs PRA Evaluation Response, SQN-0-13-072. The CIVs marked as "Not Analyzed" either did not match the generic configurations and it was not advantageous to perform a specific SQN analysis that would increase the CT times greater than the original 4 hours or generic configurations did not yield a CT greater than the original 4 hours. ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 Disposition of Existing License Amendment Requests DISPOSITION OF EXISTING LICENSE AMENDMENT REQUESTS The following License Amendment Requests are under NRC review. The following table describes the request, and its affect on the ITS conversion, and its disposition. DISPOSITION OF EXISTING LICENSE AMENDMENT REQUESTS Submittal Date Description of Change Affected ITS Submittal Sections/ Specifications Affected CTS Pages Disposition August 10, 2012 Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis, (SQN-TS-12-02) None None This is currently with the NRC for review. January 7, 2013 Sequoyah Nuclear Plant, Units 1 and 2 License Renewal None None This is currently with the NRC for review. July 3, 2013 Application to Modify Ice Condenser Technical Specifications to Address Revisions in Westinghouse Mass and Energy Release Calculation (SQN-TS-12-04) ITS: 3.6.12 Unit 1 3/4 6-26, 3/4 6-27 Unit 2 3/4 6-27, 3/4 6-28 Proposed changes are already reflected in this ITS submittal. Changes are annotated with an "A" DOC referencing the previously submitted LAR. See ITS 3.6.12 DOC A02. October 2, 2013 Sequoyah Nuclear Plant (SQN), Units 1 and 2 - Proposed Technical Specification (TS) Change, "Ultimate Heat Sink (UHS) Temperature Limitations Supporting Alternate Essential Raw Cooling Water (ERCW) Loop Alignments (TS-SQN-13-01 and 13-02)" ITS: 3.7.9 Units 1 and 2 3/4 7-14 Proposed changes are already reflected in this ITS submittal. Changes are annotated with an "A" DOC referencing the previously submitted LAR. See ITS 3.7.9 DOC A02. ENCLOSURE 8 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 Regulatory Commitments REGULATORY COMMITMENTS No. Commitments for TSTF-411/418 Due Date/Event 1 Activities that degrade the availability of the AFW system, Reactor Coolant System (RCS) pressure relief system (pressurizer PORVs and safety valves), AMSAC, or Turbine Trip should not be scheduled when a logic cabinet is unavailable. Upon Implementation 2 One complete Emergency Core Cooling System (ECCS) train that can be actuated automatically must be maintained when a logic cabinet is unavailable. Upon Implementation 3 Activities that cause analog channels to be unavailable should not be scheduled when a logic cabinet is unavailable. Upon Implementation 4 Activities on electrical systems (e.g., AC and DC power) and cooling systems (e.g., Essential Raw Cooling Water System (ERCW) and Component Cooling Water System (CCS) that support the systems or functions listed in the three commitments above (AFW, RCS pressure relief systems, AMSAC, Turbine Trip, or ECCS) should not be scheduled when a logic cabinet is unavailable. That is, one complete train of a function that supports a complete train of a function noted above must be available. Upon Implementation 5 Activities that degrade the availability of the auxiliary feedwater system, RCS pressure relief system (pressurizer PORVs and safety valves), AMSAC, or turbine trip should not be scheduled when a RTB is out of service. Upon Implementation 6 Activities that degrade other components of the RPS, including master and slave relays, and activities that cause analog channels to be unavailable should not be scheduled when a RTB is unavailable. Upon Implementation No. Commitments for TSTF-427 Due Date/Event 7 Sequoyah Unit 1 & Unit 2 will incorporate the guidance of NUMARC 93-01 Section 11, which provides guidance and details on the assessment and management of risk during maintenance. Upon Implementation 8 Sequoyah Unit 1 & Unit 2 will revise procedures to ensure that the risk assessment and management process described in NEI 04-08 is used whenever a barrier is considered unavailable and the requirements of LCO 3.0.9 are to be applied, in accordance with an overall configuration risk management program (CRMP) to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified and avoided Upon Implementation No. Commitments for TSTF-446 Due Date/Event 9 Sequoyah Unit 1 & Unit 2 will implement the capability to assess the effect on incremental large early release probability when using the extended completion times for containment isolation valves in the program for managing risk in accordance with 10 CFR 50.65(a)(4) and the plant-specific configuration risk management program. Upon Implementation No. Commitments for TSTF-493Due Date/Event10 Sequoyah will revise the UFSAR to include the methodologies used to determine the as-found and as-left tolerances for Limiting Safety Setting System (LSSS) instrument channel setpoints. Upon Implementation 11 Sequoyah will develop a monitoring program to adequately track the performance of Master Relays, Slave Relays, Logic Cabinets, Universal Logic Cards, Undervoltage Driver Cards, Safeguards Driver Cards, and Reactor Trip Breakers. (Reference Westinghouse Reports Section 3.2 and 3.5) Upon Implementation No. Commitment for ITS 3.7.12 Condition B Due Date/Event 12 Sequoyah will have guidance available describing compensatory measures to be taken in the event of an intentional or unintentional entry into ITS 3.7.12 Condition B. Upon Implementation No. Commitment for ITS 3.9.4 Reviewer's Note Due Date/Event 13 The following guidelines are included in the assessment of systems removed from service during movement irradiated fuel:

- During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the Technical Specification OPERABILITY amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

- A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure.

Upon Implementation The purpose of the "prompt methods" mentioned above are to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored." The above table identifies 11 commitments by TVA in Enclosure 8 for the SQN conversion to Improved Technical Specifications license amendment request (LAR). Any other statements in this LAR submittal are provided for informational purposes and are not considered regulatory commitments. ENCLOSURE 9 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 List of Required Final Safety Analysis Report (FSAR) Descriptions For TSTF-500 LIST OF REQUIRED FINAL SAFETY ANALYSIS REPORT (FSAR) DESCRIPTIONS FOR TSTF-500 The following table identifies FSAR descriptions for the Diesel Generator and Vital Batteries required by Sequoyah Nuclear Plant, Units 1 and Unit 2, as part of the adoption of TSTF-500, Revision 2. These changes will be included with the required implementation date in the Issuance of Amendment letter. REQUIRED FSAR DESCRIPTION DUE DATE/EVENT Sequoyah will change or verify that the FSAR: 1.Describes how a 5 percent design margin for the 125V Vitalbatteries corresponds to a 2 amp float current valueindicating that the battery is 98 percent charged.2.Describes how a 5 percent design margin for the DieselGenerator batteries corresponds to a 1 amp float current value indicating that the battery is 98 percent charged.3.States that long term battery performance is supported bymaintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 2.13 V per connectedcell and that there are 60 connected cells in the battery,which corresponds to 127.8 V at the battery terminals.4.Describes how the batteries are sized with correctionmargins that include temperature and aging and how thesemargins are maintained.5.States the minimum established design limit for batteryterminal float voltage.6.States the minimum established design limit for electrolytelevel.7.States the minimum established design limit for electrolytetemperature.8.Describes how each battery is designed with additionalcapacity above that required by the design duty cycles toallow for temperature variations and other factors.9.Describes normal DC system operation (i.e., powered fromthe battery chargers) with the batteries floating on thesystem, and a loss of normal power to the battery chargerdescribing how the DC load is automatically powered from the station batteries.Upon implementation (applies to all)

10. Describes the availability of a means to charge the Vital Batteries and a description that the battery charger is capable of being supplied power from a power source that is independent of the offsite power supply. Specification 3.8.4, Required Action A.3 11. Describes that DG tests verify that the critical protective trips that are not automatically bypassed perform their intended function. Upon implementation ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 SQN Self-Identified Issues SQN Self Identified Issues During the NRC staff's review process, the staff had multiple requests for additional information (RAIs). In order to provide responses to the RAIs, SQN staff reviewed the SQN ITS conversion numerous times. As a result of the review, several issues were identified by SQN requiring revisions to the originally submitted ITS LAR. The table below provides information concerning justifications for the required revisions, ITS sections affected by the required revisions, and the location in the original LAR where the revisions occur.

Justification Section Page # The Summary Disposition Matrix has been revised to indicate that CTS 3.9.3, Decay Time, is retained in ITS 3.9.8, Decay Time. This change ensures a fuel handling accident, involving recently irradiated fuel assemblies, does not occur. Summary Disposition Matrix Enclosure 2, Volume 1, page 30 of 37 CTS Section 2.1 Safety Limits, the Reactor Core Safety Limit Figure, 2.1-1 was originally proposed for relocation to the COLR and discussed in DOC LA01. This Figure will be retained in ITS 2.1.1 as Figure 2.1.1-1 and DOC LA01 will be deleted. This change maintains CTS. ITS 2.1.1 (Units 1 & 2) Enclosure 2, Volume 1, page 14 of 37 Enclosure 2, Volume 4, pages 5, 6, 8, 9, 11, 12, 14, 15a, 16, 17a, 22, 23, 30, and 31 of 38 Information has been added to the ISTS 3.0 Bases to clarify when SR 3.0.2 and SR 3.0.3 are applicable to Chapter 5 Specifications. ITS 3.0 Bases Enclosure 2, Volume 5, pages 57, 58, 59, 78, 79, 80, and 84 of 90 Inadvertent Omission of CTS SR number in DOC LA01 - Editorial ITS 3.1.2 (DOC LA01) Enclosure 2, Volume 6, page 44 of 356 Correction of inadvertent deletion of "INSERT 5" flag, JFD 4 indicators, and INSERT 5 heading and JFD 4 indicators on following page. - Editorial ITS 3.1.7 Bases (Units 1 & 2) Enclosure 2, Volume 6, pages 271, 272, 283 and 284 of 356 CTS 4.2.2.2.c.3 and CTS 4.2.2.2.c.4 require actions if the AFD min margin or the f2(I) min margin are < 0. Therefore, ITS SR 3.2.1.2 and SR 3.2.1.3 should verify the AFD min margin and the f2(I) min margin are 0. Associated changes will be required for JFD 4, JFD 6, and the ITS Bases. This change maintains CTS. ITS 3.2.1 (Units 1 & 2) Enclosure 2, Volume 7, pages 35, 37, 47, 49, 51, 70, and 93 of 249 SQN Self Identified Issues CTS 4.2.3.2.c.3 and CTS 4.2.3.2.c.4 require actions if the FH min margin or the f1(I) min margin are < 0. Therefore, ITS SR 3.2.2.1 and SR 3.2.2.2 should verify the FH min margin and the f1(I) min margin are 0. Associated changes will be required for DOC M01, JFD 8, and the ITS 3.2.2 Bases. This change maintains CTS. ITS 3.2.2 (Units 1 & 2) Enclosure 2, Volume 7, pages 114, 126, 128, 129, 134, 136, 137, 138, 145, 149, 160, and 164 of 249 ITS 3.2.3 Bases correction associated with the details of the resolution from CSS-007. Resolution requires one additional sentence, "This ensures that the fuel cladding integrity is maintained for these postulated accidents," be deleted (see UFSAR 15.5.3). This change maintains current licensing basis. ITS 3.2.3 Bases (Units 1 & 2) Enclosure 2, Volume 7, pages 191 and 197 of 249 CTS Section 3.2, TABLE 3.2-1, DNB Parameters, contains the specific values for RCS average temperature, pressurizer pressure and RCS total flow rate. These specific values were proposed to be relocated to the COLR, as discussed in DOC LA01, and are now being retained in ITS LCO 3.4.1. Associated changes will be required for DOC LA03, DOC M01, ITS SRs 3.4.1.1, 3.4.1.2, 3.4.1.3, and 3.4.1.4, and the ITS 3.4.1 Bases. This change maintains CTS. ITS 3.4.1 (Units 1 & 2) Enclosure 2, Volume 9, pages 6, 7, 9, 10, 11, 11a, 13, 16, 18, 19, 20, 22, 23, 24, 28, and 37 of 696 Correction to Bases References section - Editorial ITS 3.4.1 & ITS 3.4.12 (Units 1 & 2) Enclosure 2, Volume 9, pages 34, 43, 421 and 441 of 696 SQN Self Identified Issues CTS LCO 3.4.6.3 page used to markup the conversion to ITS SR 3.4.14.1 did not reflect the correct limit for RCS PIV leakage as compared to the controlled copy of the SQN TS. This change maintains CTS. ITS SR 3.4.14.1 (Units 1 & 2) Enclosure 2, Volume 9, pages 479, 484, 494, and 498 of 696 Correction of ITS SR 3.4.14.1 Surveillance Frequency to both the IST and SFCP. Associated changes will be required for the ITS 3.4.14 Bases. This change maintains CTS. ITS SR 3.4.14.1 (Units 1 & 2) Enclosure 2, Volume 9, pages 494, 498, 506, 508, 514 and 516 of 696 Correction of referenced ITS SR Number. - Editorial ITS 3.4.14 - JFD 7 Enclosure 2, Volume 9, page 500 of 696 Correction of insert for Required Actions B.1 and B.2 - Editorial ITS 3.4.14 Bases (Units 1 & 2) Enclosure 2, Volume 9, pages 505 and 513 of 696 Correction of Required Action Header for B.1.1, B.1.2, and B.2 - Editorial ITS 3.4.15 (Units 1 & 2) Bases Enclosure 2, Volume 9, pages 560 and 568 of 696 Correction of Unit 2 specific information in "Insert 2" - This change maintains CTS. ITS 4.0 (Unit 2) Enclosure 2, Volume 15, page 29 of 35 Correction of acronym from ASME to ANSI for ANSI N510-1975. This change maintains CTS. ITS 5.5.9 (Units 1 & 2) Enclosure 2, Volume 16, pages 162 and 190 of 270 Restoration of the CTS acceptance criteria for Pa in ITS 5.5.14. This change maintains CTS. ITS 5.5.14 (Unit 1 & 2) Enclosure 2, Volume 16, pages 96, 127, 128, 140, 140a, 146, 147, 168, 169, 196, 197 and 205a of 270 CTS 6.9.1.14, COLR, information contained in CTS 2.1.1, 3.3.1, and 3.9.1 was proposed for relocation to the COLR. This information will be retained in ITS. Therefore, the list of proposed information relocated to the COLR will be revised. Associated changes are required for DOC A05. This change maintains CTS. ITS 5.6.3 (Unit 1 & 2) Enclosure 2, Volume 16, pages 213, 221, 227, 233 and 241 of 270 Correction in ISTS 5.7.2 - Editorial ITS 5.7.2 (Unit 1 & 2) Enclosure 2, Volume 16, page 263 and 267 of 270 SQN Self Identified Issues Correction to status of TS-SQN-13-01 and 13 Editorial Enclosure 6 Enclosure 6, page E6-1}}

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