L-2023-063, Enclosure 2, Volume 1: Turkey Point Nuclear Plant, Units 3 and 4, Improved Technical Specifications Conversion, Application of Selection Criteria to Technical Specifications, Revision 2, Volume 1
| ML23151A439 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/24/2023 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23151A435 | List:
|
| References | |
| L-2023-063 | |
| Download: ML23151A439 (1) | |
Text
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Enclosure 2 Dockets 50-250 and 50-251 Volume 1 ENCLOSURE 2 VOLUME 1 TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION APPLICATION OF SELECTION CRITERIA TO THE TECHNICAL SPECIFICATIONS REVISION 2 R2
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 2 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS CONTENTS Page
1.0 INTRODUCTION
........................................................................................ 1 2.0 SELECTION CRITERIA .............................................................................. 1 3.0 PRA INSIGHTS ........................................................................................... 2 4.0 RESULTS OF APPLICATION OF SELECTION CRITERIA ........................ 6
5.0 REFERENCES
............................................................................................ 6 ATTACHMENTS ATTACHMENT 1
SUMMARY
DISPOSITION MATRIX FOR TURKEY POINT NUCLEAR POWER PLANT UNIT 3 AND UNIT 4 APPENDIX APPENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION
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- 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 Turkey Point Nuclear Generating Station Unit 3 and Unit 4 (hereinafter PTN Units 3 and 4). R2 Florida Power & Light Company (hereinafter FPL) 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, 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 as revised in NUREG-1431, Revision 5.0, "Standard Technical Specifications, Westinghouse Plants" (Reference 2), and applied the criteria to each of the current PTN 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 PTN.
- 2. SELECTION CRITERIA FPL 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 FPL, 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
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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, (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.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 5 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS 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, 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,
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- 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 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 Objects Reference 3 includes a statement that the 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 or similar processes subject to NRC review. 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
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 7 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS 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 FPL for the applicable Turkey Point (PTN) Specifications to be relocated.
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 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
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- 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.
- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the PTN Units 3 and 4 Technical R2 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.
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- 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 Specification, Westinghouse Plants," Revision 5.0
- 3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132)
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SUMMARY
DISPOSITION MATRIX FOR PTN UNITS 3 AND 4 R2
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 1.0 Use and Applications This section provides definitions for several defined terms used throughout the remainder of the Technical Specifications (TSs). The definitions are provided to identify the meaning of certain terms. As 1.1 Definitions such, direct application of the selection 1.0 Definitions Yes-NA 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 the Improved Technical Specifications (ITS).
NA NA 1.2 Logical Connectors NA NA NA 1.3 Completion Times NA NA NA 1.4 Frequency NA Safety Limits and 2.0 Limiting Safety System Settings 2.1 Safety Limits Application of Technical Specification selection criteria is not appropriate. R2 2.1.1 Reactor Core 2.1.1 Reactor Core SLs Yes-NA However, Safety Limits are included in the ITS as required by 10 CFR 50.36.
Reactor Coolant System 2.1.2 Reactor Coolant R2 2.1.2 Yes-NA Same as above.
Pressure System Pressure SL Limiting Safety System 2.2 Settings The Reactor Trip System (RTS) and R2 Limiting Safety System Settings (LSSS) are included as part of the RPS Reactor Trip System 3.3.1 Reactor Trip instrumentation Specification, which is R2 2.2.1 Instrumentation System (RTS) Yes-3 retained since the functions either actuate Setpoints Instrumentation to mitigate the consequences of DBAs and transients or are retained as directed by the NRC as the functions are part of RPS.
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION Limiting Conditions for 3.0 Operation This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of Operational Mode Limiting Conditions for Operations 3.0.1 applicability for LCO 3.0.1 Yes-NA (LCOs)and Surveillance Requirements requirements (SRs). As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements are retained in the ITS, as modified consistent with NUREG-1431.
Compliance with the 3.0.2 LCO 3.0.2 Yes-NA Same as above.
Specifications Generic Actions for 3.0.3 LCO 3.0.3 Yes-NA Same as above.
noncompliance Entry into Operational 3.0.4 LCO 3.0.4 Yes-NA Same as above.
Mode restrictions This Specification is being retained in the LCO and ACTION ITS to clarify the application of Turkey 3.0.5 Application for Combined LCO 3.0.10 Yes-NA Point Unit 3 and Unit 4 (PTN) Combined TS TS for TS LCO and Actions.
This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of LCOs and SRs. As such, direct 3.0.6 Action Exceptions LCO 3.0.5 Yes-NA application of the Technical Specification selection criteria is not appropriate.
However, the general requirements are retained in the ITS, as modified consistent with NUREG-1431.
Support/Supported 3.0.7 LCO 3.0.6 Yes-NA Same as above Systems NA LCO 3.0.7 Test Exceptions NA LCO 3.0.8 Snubbers
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION NA LCO 3.0.9 Barriers Surveillance 4.0 4.0 SR Applicability Requirements This Specification provides generic guidance applicable to one or more Specifications. The information is Operational Mode provided to facilitate understanding of applicability for LCOs and SRs. As such, direct 4.0.1 SR 3.0.1 Yes-NA surveillance application of the Technical Specification requirements selection criteria is not appropriate.
However, the general requirements are retained in the ITS, as modified consistent with NUREG-1431, Revision 5.
4.0.2 Time of performance SR 3.0.2 Yes-NA Same as above.
Compliance with 4.0.3 surveillance SR 3.0.3 Yes-NA Same as above.
requirements Entry into Operational 4.0.4 SR 3.0.4 Yes-NA Same as above.
Modes This requirement was deleted in Current 4.0.5 Deleted NA NA Technical Specification (CTS)
Amendments 281 and 275.
This Specification is being retained in the Application of SRs with 4.0.6 SR 3.0.5 Yes-NA ITS to clarify the application of PTN Combined TS Combined TS for TS SRs.
Reactivity Control 3.1 Reactivity Control 3/4.1 Systems Systems 3.1.1 Shutdown Margin (SDM) 3.1.2 Core Reactivity Shutdown Margin - Tavg 3.1.1.1 3.1.5 Shutdown Bank Yes-2 Greater Than 200°F Insertion Limits 3.1.6 Control Bank Insertion Limits SDM Tavg Less Than or 3.1.1.2 3.1.1 SDM Yes-2 Equal to 200°F
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3.1.3 Moderator Moderator Temperature 3.1.1.3 Temperature Coefficient Yes-2 Coefficient (MTC) 3.4.2 RCS Minimum Minimum Temperature 3.1.1.4 Temperature for Yes-2 for Criticality Criticality 3.1.2.1 Flow Path - Shutdown NA No Relocate. See Appendix A.
3.1.2.2 Flow Paths - Operating NA No Relocate. See Appendix A.
Charging Pumps -
3.1.2.3 NA No Relocate. See Appendix A.
Operating Borated Water Source -
3.1.2.4 NA No Relocate. See Appendix A.
Shutdown Borated Water Sources -
3.1.2.5 NA No Relocate. See Appendix A.
Operating 3.1.4 Rod Group 3.1.3.1 Group Height Yes-2 Alignment Limits Position Indication 3.1.7 Rod Position 3.1.3.2 Yes-2 Systems - Operating Indication Position Indication 3.1.3.3 NA No Relocate. See Appendix A.
System - Shutdown SR 3.1.4.3 Verify rod 3.1.3.4 Rod Drop Time Yes-2 drop time Shutdown Rod Insertion 3.1.5 Shutdown Bank 3.1.3.5 Yes-2 Limit Insertion Limit Control Rod Insertion 3.1.6 Control Bank 3.1.3.6 Yes-2 Limits Insertion Limits Power Distribution 3.2 Power Distribution 3/4.2 Limits Limits 3.2.3 Axial Flux 3.2.1 Axial Flux Difference Yes-2 Difference (AFD)
Heat flux Hot Channel 3.2.1 Heat Flux Hot 3.2.2 Yes-2 Factor FQ(Z) Channel Factor (FQ(Z) 3.2.3 3.2.2 Nuclear Enthalpy Yes-2 Nuclear Enthalpy Rise Rise Hot Channel Factor
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION Hot Channel Factor N (F H
)
Quadrant Power Tilt 3.2.4 Quadrant Power 3.2.4 Yes-2 Ratio Tilt Ratio (QPTR) 3.4.1 RCS Pressure, Temperature, and Flow 3.2.5 DNB Parameters Yes-2 Departure from Nucleate Boiling (DNB) Limits 3/4.3 Instrumentation 3.3 Instrumentation 3.3.1 Reactor Trip Reactor Trip System 3.3.1 System (RTS) Yes-3 Instrumentation Instrumentation 3.3.2 Engineered Safety Features Actuation System (ESFAS)
Instrumentation 3.3.4 Control Room Emergency Ventilation System (CREVS)
Actuation 3.3.2 ESFAS Instrumentation Yes-3 3.3.5 Loss of Power Emergency Diesel Generator (EDG) Start Instrumentation 3.3.6 Containment Ventilation Isolation Instrumentation 3.3.6 Containment Ventilation Isolation Radiation Monitoring Instrumentation R2 3.3.3.1 Yes-1/3 Instrumentation 3.4.15 RCS Leakage Detection Instrumentation Movable Incore 3.3.3.2 NA No Relocate. See Appendix A.
Detectors
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION Accident Monitoring Instrumentation 3.3.3 Post Accident (Instruments 1, 2, 3, 4, 5, 3.3.3.3 Monitoring (PAM) Yes-3 6, 13, 14, 15, 16, 17, 20, Instrumentation 21, and 22 are being retained.)
Accident Monitoring Instrumentation 3.3.3.3 (Instruments 7, 8, 9, 10, NA No Relocate. See Appendix A 11, and 12 are being relocated.
3.4 Reactor Coolant 3.4.4 Reactor Coolant System System Reactor Coolant Loops and Coolant Circulation - 3.4.4 RCS Loops -
3.4.1.1 Yes-2 Startup and Power MODES 1 and 2 Operation 3.4.5 RCS Loops -
3.4.1.2 RCS - Hot Standby Yes-3 MODE 3 3.4.6 RCS Loops -
3.4.1.3 RCS - Hot Shutdown Yes-4 MODE 4 RCS Cold Shutdown - 3.4.7 RCS Loops -
3.4.1.4.1 Yes-4 Loops Filled MODE 5, Loops Filled 3.4.8 RCS Loops -
Cold Shutdown - Loops 3.4.1.4.2 MODE 5, Loops Not Yes-4 Not Filled Filled Safety Valves - 3.4.10 Pressure Safety R2 3.4.2.1 Yes-3 Shutdown Valves Safety Valves - 3.4.10 Pressure Safety 3.4.2.2 Yes-3 Operating Valves R2 3.4.3 Pressurizer 3.4.9 Pressurizer Yes-2 3.4.11 Pressurizer Power 3.4.4 Relief Valves Operated Relief Valves Yes-3 (PORVs) Block Valves Steam Generator (SG) 3.4.17 Steam Generator 3.4.5 Yes-2 SG Tube Integrity is 3.4.20 in the NUREG Tube Integrity (SG) Tube Integrity
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3.4.15 RCS Leakage RCS Leakage Detection 3.4.6.1 Detection Yes-1 Systems Instrumentation 3.4.13 RCS Operational Leakage RCS Operational 3.4.6.2 3.4.14 RCS Pressure Yes-2 Leakage R2 Isolation Valve (PIV)
Leakage 3.4.7 Deleted NA NA Deleted in CTS 3.4.16 RCS Specific 3.4.8 Specific Activity Yes-2 Activity RCS Pressure 3.4.3 RCS Pressure and 3.4.9.1 Yes-2 Temperature Limits Temperature (P/T) Limits 3.4.9.2 Pressurizer Temperature NA No Relocate. See Appendix A.
Overpressure Mitigating 3.4.12 Overpressure 3.4.9.3 Yes-2 Systems Mitigating System 3.4.10 Deleted in CTS NA NA Reactor Coolant System 3.4.11 NA No Relocate. See Appendix A.
Vents 3.5 Emergency Core 3/4.5 ECCS Cooling Systems (ECCS) 3.5.1 Accumulators 3.5.1 Accumulators Yes-3 3.4.14 RCS Pressure R2 ECCS Subsystems - Tavg Isolation Valve (PIV) 3.5.2 Yes-3 350 °F Leakage 3.5.2 ECCS - Operating ECCS Subsystem -
3.5.3 3.5.3 ECCS - Shutdown Yes-3 Tavg < 350 °F Refueling Water Storage 3.5.4 Refueling Water 3.5.4 Yes-3 Tank Storage Tank (RWST)
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3/4.6 3.6 Containment Containment Systems Systems 3.6.1.1 Containment Integrity 3.6.1 Containment Yes-3 3.6.1.2 Containment Leakage 3.6.1 Containment Yes-3 3.6.2 Containment Air 3.6.1.3 Containment Air Locks Yes-3 Locks Containment Internal 3.6.4 Containment 3.6.1.4 Yes-2 Pressure Pressure Containment Air 3.6.5 Containment Air 3.6.1.5 Yes-2 Temperature Temperature Containment Structural 3.6.1.6 3.6.1 Containment Yes-3 Integrity Containment Ventilation 3.6.3 Containment 3.6.1.7 Yes-3 System Isolation Valves Containment Spray 3.6.6 Containment Spray 3.6.2.1 Yes-3 System and Cooling Systems Emergency Containment 3.6.6 Containment Spray 3.6.2.2 Yes-3 Cooling System and Cooling Systems Recirculation pH Control 3.6.7 Recirculation pH 3.6.2.3 Yes-3 System Control System 3.6.3 Containment Containment Isolation Isolation valves 3.6.4 Yes-3 R2 Valves 3.9.4 Containment Penetrations 3/4.7 Plant Systems 3.7 Plant Systems 3.7.1 Main Steam Safety 3.7.1.1 Safety Valves Yes-3 Valves (MSSVs)
Auxiliary feedwater 3.7.5 Auxiliary Feedwater 3.7.1.2 Yes-3 System (AFW) System Condensate Storage 3.7.6 Condensate 3.7.1.3 Yes-2/3 Tank Storage Tank (CST)
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3.7.4 Secondary Specific Secondary Specific Activity is 3.7.18 in 3.7.1.4 Specific Activity Yes-2 Activity NUREG.
Main Steam Isolation 3.7.2 Main Steam 3.7.1.5 Yes-2 Valves Isolation Valves (MSIVs) 3.7.1.6 Deleted NA NA Deleted in Amendments 282 and 276 3.7.3 Feedwater Isolation Valves (FIVs) and 3.7.1.7 Feedwater Isolation Yes-3 Feedwater Control Valves (FCVs) 3.7.7 Component Component Cooling 3.7.2 Cooling Water (CCW) Yes-3 Water System System 3.7.8 Intake Cooling Intake Cooling Water 3.7.3 Water (ICW) System Yes-3 System (ICWS) 3.7.9 Ultimate Heat Sink 3.7.4 Ultimate Heat Sink Yes-3 (UHS) 3.7.10 Control Room Emergency Ventilation Control Room System (CREVS) 3.7.5 Emergency Ventilation 3.7.11 Control Room Yes-3 System Emergency Air R2 Temperature Control System (CREATCS)
General requirements associated with snubbers that affect Technical Specification systems, trains, subsystems, and components are included in ITS LCO 3.0.8, as 3.7.6 Snubbers NA No modified consistent with NUREG-1431.
See technical change discussions related to CTS 3/4.7.6 in the Discussion of Changes for ITS Sections 3.0 and 3.7.
Sealed Source 3.7.7 NA No Relocate. See Appendix A.
Contamination R2
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 20 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Attachment 1 - Summary Disposition Matrix For PTN RETAINED/
CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3.8 Electrical Power 3/4.8 Electrical Power Systems Systems 3.8.1 AC Sources -
Operating 3.8.1.1 AC Sources - Operating Yes-3 3.8.3 Diesel Fuel Oil, R2 Lube Oil, and Starting Air 3.8.2 AC Sources -
Shutdown 3.8.1.2 AC Sources - Shutdown Yes-3 3.8.3 Diesel Fuel Oil, R2 Lube Oil, and Starting Air 3.8.4 DC Sources -
Operating 3.8.6 Battery Parameters 3.8.2.1 DC Sources - Operating Yes-3 5.5.14 Battery R2 Maintenance and Monitoring Program 3.8.5 DC Sources -
3.8.2.2 DC Sources - Shutdown Shutdown Yes-3 3.8.6 Battery Parameters R2 3.8.7 Inverters -
R2 Onsite Power Distribution Operating 3.8.3.1 Yes-3 Added separate Inverter LCO to PTN ITS
- Operating 3.8.9 Distribution Systems - Operating 3.8.8 Inverters - R2 Onsite Power Distribution Shutdown 3.8.3.2 Yes-3 Added separate Inverter LCO to PTN ITS
- Shutdown 3.8.10 Distribution Systems - Shutdown 3/4.9 Refueling Operations 3.9 Refueling Operations 3.9.1 Boron 3.9.1 Boron concentration Yes-2 Concentration 3.9.3 Nuclear 3.9.2 Instrumentation Yes-3 Instrumentation 3.9.3 Decay Time NA NA LA to the TRM Containment Building 3.9.4 Containment 3.9.4 Yes-3 Penetrations Penetrations
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 21 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Attachment 1 - Summary Disposition Matrix For PTN RETAINED/
CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION 3.9.5 Deleted NA NA Deleted in CTS 3.9.6 Deleted NA NA Deleted in CTS 3.9.7 Deleted NA NA Deleted in CTS 3.9.5 Residual Heat Residual Heat Removal Removal and Coolant 3.9.8.1 and Coolant Circulation - Yes-3 Circulation - High Water High Water Level Level 3.9.6 Residual Heat Refueling Operations - Removal and Coolant 3.9.8.2 Yes-3 Low Water Level Circulation - Low Water Level Containment Ventilation 3.9.4 Containment 3.9.9 Yes-3 Isolation System Penetrations Refueling Cavity Water 3.9.2 Refueling Cavity Refueling Cavity Water Level is 3.9.7 in 3.9.10 Yes-3 Level Water Level the NUREG.
Water Level Storage 3.7.12 Fuel Storage Pool Fuel Storage Pool Water Level is 3.7.15 in R2 3.9.11 Yes-3 Pool Water Level NUREG.
3.9.12 Deleted NA NA Deleted in CTS 3.3.6 Containment 3.9.13 Radiation Monitoring Ventilation Isolation Yes-3 Instrumentation 3.7.13 Fuel Storage Pool Fuel Storage Pool Boron Concentration is R2 Boron Concentration 3.9.14 Spent Fuel Storage Yes-3 3.7.16 in NUREG 3.7.14 Spent Fuel Spent Fuel Storage is 3.7.17 in NUREG Storage 3/4.10 Special Test Exceptions Special Test Exceptions Delete - See Discussion of Changes for 3.10.1 NA No Shutdown Margin CTS 3/4.10.1.
Special Test Exceptions Group Height, Insertion, Delete - Discussion of Changes for CTS 3.10.2 NA No and Power Distribution 3/4.10.2.
Limits
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CURRENT CURRENT NEW TS (ITS) CRITERION NOTES TS TITLE NUMBER FOR INCLUSION Although this Specification does not meet any Technical Specification selection Special Test Exceptions 3.1.8 PHYSICS TESTS criteria, it has been retained to provide 3.10.3 Yes-3 Physics Tests Exceptions - MODE 2 flexibility to perform certain operations by appropriately modifying requirements of other LCOs.
3.10.4 NOT USED NA NA Special Test Exceptions See Discussion of Changes for CTS 3.10.5 Position Indication NA No 3.10.5.
System - Shutdown 3.7.14 Spent Fuel 5.0 DESIGN FEATURES Storage Yes-NA 4.0 Design Features R2 5.0 Administrative 6.0 Administrative Controls Yes-NA Controls
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 23 of 35 APPENDIX A JUISTIFICATION FOR SPECIFICATION RELOCATION
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 24 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.1.2.1, FLOW PATHS - SHUTDOWN CTS 3.1.2.2, FLOW PATHS - OPERATING CTS 3.1.2.3, CHARGING PUMPS - OPERATING CTS 3.1.2.4, BORATED WATER SOURCE - SHUTDOWN CTS 3.1.2.5, BORATED WATER SOURCE - SHUTDOWN Discussion:
Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) 3.1.2.1 provides the requirements for the minimum boron injection flow paths (one) during shutdown (Modes 5 and 6). CTS 3.1.2.2 provides the requirements for the minimum boron injection flow paths (2) during operation Modes 1 - 4. CTS 3.1.2.3 provides the requirement to have two charging pumps available during Modes 1 - 4 as the motive means to transfer the boron inventory to the Reactor Coolant System (RCS) during normal operation. CTS 3.1.2.4 requires, as a minimum, one borated water source (Boric Acid Storage System or Refueling Water Storage Tank) to be operable during Modes 5 and 6. CTS 3.1.2.5 requires both borated water sources (Boric Acid Storage System and Refueling Water Storage Tank) to be operable during Modes 1 - 4.
The components associated with the boration system Technical Specifications (TSs) provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin (SDM) during normal operations. To accomplish this functional requirement, the current boration system TSs require a source(s) of borated water, one or more flow paths to inject borated water into the RCS and Charging Pumps to provide the necessary charging head.
The boration systems are not assumed to be operable to mitigate the consequences of a Design Basis Accident (DBA) or transient. In the case of a malfunction of a component in the boration systems which causes a boron dilution event, the automatic response, or that required by the operator, is to close the appropriate valves in the reactor makeup system. The automatic plant response to a boron dilution event also includes automatic control rod assembly movement and reactor trip features to ensure shutdown margin is maintained. The boration capabilities of the boration systems are not assumed to mitigate the boron dilution event.
Comparison to Selection Criteria:
- 1. The boration systems do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The boration systems are not a process variable, design feature, or operating restriction
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- 3. The boration systems are not 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.
- 4. The function of injecting borated water to maintain shutdown margin is not risk significant. Operational experience has shown that the boration management system is not a constraint of prime importance in the mitigation of any accident or transient that results in challenging public health and safety. Therefore, the RCS boration management system functions to control boron concentration and maintain shutdown margin do not represent structures, systems, or components which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
==
Conclusion:==
Since the selection criteria have not been satisfied, the boration system Limiting Conditions for Operation (LCOs) and Surveillances, may be relocated to licensee-controlled documents outside the TSs. Operability requirements for Emergency Boration and Residual Heat Removal (RHR) pumps that support core RHR capability and minimum boration requirements during plant shutdown, are retained in separate TSs.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 26 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.1.3.3, POSITION INDICATION SYSTEM - SHUTDOWN Discussion:
CTS 3.1.3.3 provides the requirements for the group step counter demand position indicator to be operable and capable of determining within +/- 2 steps the demand position for each shutdown and control rod not fully inserted in Modes 3, 4 and 5 with the reactor trip breakers in the closed position.
Rod position indication 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. In Modes 3, 4, and 5, SDM is required per LCO 3.1.1 which references the Core Operating Limits Report (COLR). The COLR requires sufficient reactivity margin to ensure fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences.
This sufficient reactivity margin takes into account rod positions with the single rod cluster assembly of the highest reactivity worth fully withdrawn. 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.
Comparison to Selection Criteria:
- 1. Control Rod Position Indications in Modes 3, 4, and 5 do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. Control Rod Position Indications in Modes 3, 4, and 5 are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This TS specifies limits on process variables consistent with the structural analysis results. These limits, however, do not reflect initial condition assumptions in the DBA.
- 3. Control Rod Position Indications in Modes 3, 4, and 5 are not 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.
- 4. Control Rod Position Indications in Modes 3, 4, and 5 were found to be non-significant risk contributor to core damage frequency and offsite releases. These indications are not structures, systems, or components that operating experience or probabilistic safety assessment has shown to be significant to the public health and safety.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 27 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation
==
Conclusion:==
Since the selection criteria have not been satisfied, the Modes 3, 4, and 5 Control Rod Position indications in LCO and Surveillances may be relocated to licensee-controlled documents outside the TSs. Position Indication requirements in Modes 1 and 2 are required by LCO 3.1.7 to ensure the initial conditions of the Safety Analyses are maintained.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 28 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.3.3.2, MOVABLE INCORE DETECTORS Discussion:
CTS 3.3.3.2 provides the requirements for Movable Incore Detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The operability of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
This requirement and the associated Surveillance Requirements (SRs) bear no relation to the conditions or limitations that are necessary to ensure safe reactor operation. While the incores can provide monitoring capability, the detectors are mainly utilized to recalibrate the excore detectors and do not provide input to any trip function. The excores are credited and utilized to provide input to the Reactor Trip System (RTS).
Comparison to Selection Criteria:
- 1. Movable Incore Detectors do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. Movable Incore Detectors are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This TS specifies limits on process variables consistent with the structural analysis results.
These limits, however, do not reflect initial condition assumptions in the DBA.
- 3. Movable Incore Detectors are not 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.
- 4. Movable Incore Detectors were found to be non-significant risk contributor to core damage frequency and offsite releases. These indications are not structures, systems, or components that operating experience or probabilistic safety assessment has shown to be significant to the public health and safety.
==
Conclusion:==
Since the selection criteria have not been satisfied, the Movable Incore Detectors LCO and Surveillances, may be relocated to licensee-controlled documents outside the TSs. The
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 29 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation movable incores do not provide input to any trip system. The excores are credited and utilized in the RPS.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 30 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.3.3.3, Table 3.3-5, Accident Monitoring Instrumentation, Instruments 7 through 12 Discussion:
Accident monitoring instrumentation ensures sufficient information is available following an accident to allow an operator to verify the response of automatic safety systems, and to take preplanned manual actions to accomplish a safe shutdown of the plant. The NRC position on application of the screening criteria to post-accident monitoring instrumentation is documented in a letter dated May 9, 1988 from T.E. Murley (NRC) to the CE Owners Group (Reference 2). Regulatory Guide (RG) 1.97, Type A variables provide primary information; i.e., information that is essential for the direct accomplishment of the specified manual actions (including long term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for DBAs or transients. Additionally, it could not be confirmed that RG 1.97, non-Type A Category 1 variables are not of prime importance in limiting risk. Therefore, the NRC position is that the post-accident monitoring instrumentation list should contain Type A instruments and non-Type A Category 1 instruments specified in the plant's safety evaluation report (SER) on RG 1.97. Accordingly, this position has been applied to the PTN RG 1.97 instruments.
A review of the PTN UFSAR and the NRC Regulatory Guide 1.97 Safety Evaluation for PTN shows that the following CTS Table 3.3-5 Instruments do not meet Category 1 or Type A requirements:
CTS Table 3.3-5 Instrument 7 Auxiliary Feedwater Flow Rate Instrument 8 RCS Subcooling Margin Monitor Instrument 9 PORV Position Indicator (Primary Detector)
Instrument 10 PORV Block Valve Position Indicator Instrument 11 Safety Valve Position Indicator (Primary Detector)
Instrument 12 Containment Sump Level (Narrow Range)
Comparison to Selection Criteria:
- 1. These instruments are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The monitored parameters are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.
- 3. These instruments are not structures, systems, or components that are part of the primary success path or which function or actuate to mitigate a DBA or transient that
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 31 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. These post-accident monitoring instrument functions are not addressed in the PTN PRA and do not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
==
Conclusion:==
Since the selection criteria have not been satisfied for instruments which do not meet RG 1.97 Type A or non-Type A, Category 1 variable requirements, the requirements of these instruments may be relocated to a licensee controlled document outside the Technical Specifications. Post-accident monitoring instruments that monitor RG 1.97 Type A variables and non-Type A, Category 1 variables are retained in the Technical Specifications.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 32 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.4.9.2, PRESSURIZER TEMPERATURE Discussion:
CTS 3.4.9.2 provides for the maximum cooldown and heatup temperatures per hour (shall not exceed 200 °F/hr and 100 °F/hr, respectively) for the Pressurizer and the maximum spray water temperature differential (> 320 °F). The limits meet the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G. These limitations are consistent with structural analysis results. However, these limits are not initial condition assumptions of a DBA or transient. These limits represent operating restrictions and Criterion 2 includes operating restrictions. However, it should be noted that in the Final Policy Station, the Criterion 2 discussion specified only those operating restrictions required to preclude unanalyzed accidents and transients be included in TSs. This Specification does not meet the criteria for retention in the Improved Technical Specifications (ITS).
Comparison to Selection Criteria:
- 1. The pressurizer temperature limits are not installed instrumentation that are used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The pressurizer temperature limits are not a process variable, design feature, or operating restriction that is 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 pressurizer temperature limits are not a structure, system, or component that is part of 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 pressurizer temperature limits are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The pressurizer temperature limits were found to be a non-significant risk contributor to core damage frequency and offsite releases.
==
Conclusion:==
Since the selection criteria have not been satisfied, the Pressurizer Temperature limits LCO, and Surveillances may be relocated to licensee-controlled documents outside the TSs. The Pressurizer temperature limits Specification will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled by the provisions of 10 CFR 50.59.
This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 33 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.4.11, REACTOR COOLANT SYSTEM VENTS Discussion:
CTS 3.4.11 requires at least one Reactor Coolant System (RCS) vent path consisting of at least two vent valves in series powered from emergency busses to be OPERABLE and closed at the reactor vessel head and pressurizer steam space. These vents are provided to exhaust non-condensable gasses and or steam from the RCS that could inhibit natural circulation core cooling. Natural circulation is an alternate method of heat removal that is provided for in the Technical Specifications by requiring certain levels in the Steam Generators (when required) and certain water level above the reactor vessel flange during refueling. The RCS vents help facilitate natural circulation and are not included in the Improved Technical Specifications (ITS).
This Specification does not meet the criteria for retention in the ITS.
Comparison to Selection Criteria:
- 1. The RCS Vents 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.
R2
- 2. The RCS Vents are not a process variable, design feature, or operating restriction that is 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 RCS Vents are not a structure, system, or component that is part of 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 RCS Vents are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The pressurizer temperature limits were found to be a non-significant risk contributor to core damage frequency and offsite releases.
==
Conclusion:==
Since the selection criteria have not been satisfied, the RCS Vents LCO, and Surveillances may be relocated to licensee-controlled documents outside the TSs. The RCS Vents Specification will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 34 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation CTS 3.7.7, SEALED SOURCE CONTAMINATION Discussion:
CTS 3.7.7 provides the requirements that each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005 microCurie of removable contamination.
The limitations on removable contamination for sources requiring leak testing, including alpha emitters, are based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.
This requirement and the associated SRs bear no relation to the conditions or limitations that are necessary to ensure safe reactor operation.
Comparison to Selection Criteria:
- 1. Sealed Source Contamination limitations do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. Sealed Source Contamination limitations are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This TS specifies limits on process variables consistent with the structural analysis results. These limits, however, do not reflect initial condition assumptions in the DBA.
- 3. Sealed Source Contamination limitations are not 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.
- 4. Sealed Source Contamination limitations were found to be non-significant risk contributor to core damage frequency and offsite releases. These indications are not structures, systems, or components that operating experience or probabilistic safety assessment has shown to be significant to the public health and safety.
==
Conclusion:==
Since the selection criteria have not been satisfied, the Sealed Source Contamination LCO and Surveillances, may be relocated to licensee-controlled documents outside the TSs.
Turkey Point Nuclear Plant, Units 3 and 4 L-2023-063 Dockets 50-250 and 50-251 Enclosure 2, Volume 1 Page 35 of 35 APPLICATION OF SELECTION CRITERIA TO THE TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Requirements associated with the sealed sources are governed by 10 CFR Part 70.
Compliance with applicable portions of 10 CFR Part 70 is required by the operating licenses of PTN Units 3 and 4.