ML18038B754

Proposed Conversion from Current TSs to Improved STS Consistent w/NUREG-1433,rev 1

ML18038B754
Person / Time
Site:	Browns Ferry
Issue date:	09/06/1996
From:	TENNESSEE VALLEY AUTHORITY
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RTR-NUREG-1433 NUDOCS 9609190176
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[[:#Wiki_filter:BR()WNS FR~V NUCLEAR PLANT Ill'W'" tel t~F Date c of Utr Regulator Docket File Enclosure I Volume 1 9609'i90i76 '760906 PDR ADOCK 05000259 P PDR CONTENTS BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA 1.INTRODUCTION 2.SCREENING CRITERIA 3.PROBABILISTIC RISK ASSESSMENT INSIGHTS 4.RESULTS OF APPLICATION OF SCREENING CRITERIA 5.REFERENCES ATTACHMENT

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS 1, 2, AND 3 APPENDICES t A.JUSTIFICATION FOR SPECIFICATION RELOCATION B.BFN SPECIFIC RISK SIGNIFICANT EVALUATION BROMNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA INTRODUCTION The purpose of this document is to confirm the results of the BWR Owners Group application of the Technical Specification screening criteria on a plant specific basis for Browns Ferry (BFN)Units 1, 2 5 3.TVA has reviewed the application of the screening criteria to each of the Technical Specifications utilized in BWROG report NED0-31466,"Technical Specification Screening Criteria Application and Risk Assessment," (Reference 1)including Supplement 1 (Reference 1), NUREG 1433, Revision 1, Standard Technical Specifications, General Electric Plants BWR/4," and applied the criteria to each of the current BFN Units 1, 2 8 3 Technical Specifications. Additionally, in accordance with the NRC guidance, this confirmation of the application of screening criteria to BFN Units 1, 2, 5 3 includes confirming the risk insights from Probabilistic Risk Assessment (PRA)evaluations, provided in the Reference 1, as applicable to BFN Units 1, 2, E 3.SCREENING CRITERIA TVA used the screening criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 22, 1993 to develop the results contained in the attached matrix.Probabilistic Risk Assessment (PRA)insights as used in the BWROG submittal were used, confirmed by TVA, and are discussed in the next section of this report.The screening criteria and discussion provided in the NRC Final Policy statement 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). Page 1 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA 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 DBA 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 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 (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 DBA 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 DBA or transient analyses 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 DBA 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) 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 DBA 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 DBA or Transient should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequences of the DBA or Transient. Safety sequence analyses or their equivalent have been performed in recent years and provide a method of Page 2 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA 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 DBA and transient analyses, as prese~ted in Chapters 6 and 15 of the plant's FSAR (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 DBAs 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 PSA 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 PSA's have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report DBA or transient analyses.It is the intent of this criterion that those requirements that PSA or operating experience exposes as significant to public health and safety, consistent with the Commission's Safety Goal and Severe Accident Page 3 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA Policies, be retained or included in the Technical Specifications. The Commission expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PSA or risk survey and any available literature on risk insights and PSAs.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 PSAs in evaluating Technical Specifications related submittal. Further, as a part of the Commissions 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. PROBABILISTIC RISK ASSESSMENT INSIGHTS Introduction and Ob'ectives The Final Policy Statement includes a statement that NRC expects licensees to utilize the available literature on risk insights 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 for relocation to other plant controlled documents will be maintained under the 10 CFR 50.59, safety evaluation review program.These specifications have been compared to a variety of Probabilistic Risk Assessment (PRA)material with two purposes: 1)to identify if a component or variable is addressed by PRA, and 2)to judge if the component or variable is risk-important. In addition, in some cases risk was judged independent of any specific PRA material.The intent of the review was to provide a supplemental screen to the deterministic criteria.Those Technical Specifications proposed to remain part of the Improved Technical Specifications were not reviewed.This review was accomplished in Reference 1 except where discussed in Appendix A,"Justification For Specification Relocation,", and has been confirmed by TVA for those Specifications to be relocated. Where Reference 1 did not review a Technical Specification against the criteria of Reference 3, TVA performed a review similar (but not identical) to that described below for Reference 1.The results of these reviews are presented in Appendix B.Page 4 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA Assum tions and A roach Briefly, the approach used in Reference 1 was the following; The risk assessment analysis evaluated the loss of function of the system or component whose LCO was being considered for relocation and qualitatively assessed the associated effect on core damage frequency and offsite releases.The assessment was based on available-literature on plant risk insights and PRAs.Table 3-1 lists the PRAs used for making the assessments and is provided at the end of this section.A detailed quantitative calculation of the core damage and offsite release effects was not performed. However, the analysis did provide an indication of the relative significance of those LCOs proposed for relocation on the likelihood or severity of the accident sequences that are commonly found to dominate plant safety risks.The following analysis steps were performed for each LCO proposed for relocation: a~b.C.List the function(s) affected by removal of the LCO item.Determine the effect of loss of the LCO item on the function(s). Identify compensating provisions, redundancy, and backups related to the loss of the LCO item.d.e.Determine the relative frequency (high, medium, and low)of the loss of the function(s) assuming the LCO item is removed from Technical Specifications and controlled by other procedures or programs.Use information from current PRAs and related analyses to establish the relative frequency. Determine the relative significance (high, medium, and low)of the loss of the function(s). Use information from current PRAs and related analyses to establish the relative significance. Page 5 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA Apply risk category criteria to establish the potential risk significance or non-significance of the LCO item.Risk categories were defined as follows: RISK CRITERIA Consequence ~Fre uenc Hicih Medium Low High Medium Low S S NS S S NS NS NS NS S=Potential Significant Risk Contributor NS=Risk Non-Significant List any comments or caveats that apply to the above assessment. The output from the above evaluation was a list of LCOs proposed for relocation that could have potential plant safety risk significance if not properly controlled by other procedures or programs.As a result these Specifications will be relocated to other plant controlled documents outside the Technical Specifications. Page 6 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA TABLE 3-1 BWR PRAs USED IN NEDO-31466 (and Supplement 1)RISK ASSESSMENT BWR 6 Standard Plant, GESSAR II, 238 Nuclear Island, BWR/5 Standard Plant Probabilistic Risk Assessment, Docket No.STN 50-447, March 1982.La Salle Count Station, NED0-31085, Probabilistic Safety Analysis, February 1988.Grand Gulf Nuclear Station, IDCOR, Technical Report 86.2GG, Verification of IPE for Grand Gulf, March 1987.Limerick, Docket Nos.50-352, 50-353, 1981,"Probabilistic Risk Assessment, Limerick Generating Station," Philadelphia Electric Company.Shoreham, Probabilistic Risk Assessment Shoreham Nuclear Power Station, Long Island Lighting Company, SAI-372-83-PA-01, June 24, 1983.Peach Bottom 2, NUREG-75/0104,"Reactor Safety Study," WASH-1400, October 1975.Millstone Point 1, NUREG/CR-3085,"Interim Reliability Evaluation Program: Analysis of the Millstone Point Unit 1 Nuclear Power Plant," January 1983.Grand Gulf, NUREG/CR-1559,"Reactor Safety Study Methodology Applications Program: Grand Gulf Pl BWR Power Plant," October 1981.NEDC-30935P,"BWR Owners'roup Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 2," June 1987.Page 7 of 8 BROWNS FERRY NUCLEAR PLANT APPLICATION OF SCREENING CRITERIA RESULTS OF APPLICATION OF SCREENING CRITERIA The screening criteria from Section 2 were applied to the BFN Units 1, 2, and 3 Technical Specifications. The attachment is a summary of that application indicating which Specifications are being retained or relocated. Discussions that document the rationale for the relocation of each Specification which failed to meet the screening criteria are provided in Appendix A.No Significant Hazards Considerations (10 CFR 50.92)evaluations for those Specifications relocated are provided with the Justification for Changes for the specific Technical Specifications. TVA will relocate those Specifications identified as not satisfying the criteria to licensee controlled documents whose changes are governed by 10 CFR 50.59.REFERENCES NEDO-31466 (and Supplement 1),"Technical Specification Screening Criteria Application and Risk Assessment," November 1987.2.3.NUREG 1433, Revision 1,"Standard Technical Specifications, General Electric Plants BWR/4," April 1995.Final Policy Statement on Technical Specifications Improvements, July 22, 1993, (58FR39132). Page 8 of 8

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUNBER TITLE BFN ISTS NUHBER STS REV.4 NUHBER NUREG 1433 HUHBER RETAINED/CRITERION FOR INCLUS ION BASIS FOR IHCLUSIOH/EXCLUSION

1.0 Definitions

1.0 Yes Definitions for selected terms used in the Technical Specifications are provided to improve understanding and ensure consistent application. Application of the Technical Specification selection criteria to these definitions is not appropriate. However, definitions for those terms that remain in the Technical Specifications following the application of the selection criteria will be retained.1/2.1.A 1/2.1.8 1/2.1.C 2.1.A.1.c 1/2.2 Safety Limit: Fuel Cladding Integrity Safety Limit: Fuel Cladding Integrity--APRH Rod Block Trip Setting Safety Limit: Reactor Coolant System Integrity 2.1 3.3.1'3.3.5.1 3.3.5.2 3.3.6.1 Relocated 2.1~2 3.3.1.1 3.4.3 3.3.6.1 2.1.1 2.1.2 2.1.4 2.2.1 3.3.2 3.3.3 3.3.5 None 2.1.3 2.2 3.4.2.1 2.1 3.3.1.1 3.3.5.1 3.3.5.2 3.3.6.1 None 2.1.2 3.3.1.1 3.4.3 3.3.6.1 Yes No Yes Application of Technical Specification selection criteria to Safety Limits and Limiting Safety System Settings (LSSS)is not appropriate. The fuel cladding integrity LSSS (with the exception of APRH Rod Blocks)are retained by their incorporation into the RPS and ECCS instrunentation Specifications because the associated Functions either actuate to mitigate consequences of Design Basis Accidents (DBAs)and transients or are retained as directed by the HRC.See Appendix A, page A-4.Application of Technical Specification selection criteria to Safety Limits and Limiting Safety System Settings (LSSS)is not appropriate. The Reactor Coolant System integrity LSSS are retained by incorporation into the RPS and safety relief valve Specifications because the associated components function to mitigate the consequences of events that would result in overpressurization of the RCS.1.0~C Limiting Condition for Operation (LCO)Applicability LCO 3.0.3 3.8.'I 3.0.3 3/4.B.1.1 LCO 3.0.3 3.8.1 Yes This Specification provides generic guidance applicable to one or more Specifications to facilitate understanding of LCOs.As such, direct application of the Technical Specification selection criteria is not appropriate. The general requirements of 1.0.C are retained in the Technical Specifications consistent with NUREG-1433.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 1.0.LL TITLE Surveillance Requirement (SR)Applicability BFH ISTS NUMBER SR 3.0.1 SR 3.0.2 SR 3.0.3 STS REV.4 NUMBER 4.0.1 4.0.2 4.0.3 NUREG 1433 NUMBER SR 3.0.1 SR 3.0.2 SR 3.0.3 RETAIHEO/CRITERION FOR INCLUSION Yes BASIS FOR INCLUSIOH/EXCLUSIOH This Specification provides generic guidance applicable to one or more Specifications to facilitate understanding of SRs.As such, direct application of the Technical Specification selection criteria is not appropriate. The general requirements of 1.0.LL are retained in the Technical Specifications consistent with NUREG-1433. 3/4.1.A Reactor Protection System: Instrwentation that Initiate a Reactor Scram (Instrwents in Table 3.1.1 and associated SRs in Table 4.1~1 and 4.1.2)3.3~1~1 3/4.3.1 3.3.1.1 Yes-3, 4 All Functions retained (with exception listed below)because the various Functions: 1)actuate to mitigate consequences of OBAs and/or transients; or, 2)are considered risk significant and retained in accordance with the NRC Final Policy Statement on Technical Specification Improvements; or, 3)are part of the RPS/Reactor Scram Function;or, 4)provide an anticipatory scram to ensure the scram discharge voiwe and thus RPS remains operable.Table 3/4.'I.A Turbine First Stage Permissive Relocated None Hone No See RPS Instrwentation Justification for Change (LAS for ISTS 3.3.F 1)-3/4.1.B Reactor Protection System Power Supply 3.3.8.2 3/4.8.4.4 3.3.8.2 Yes-3 Provides protection for the RPS bus powered instrwentation against unacceptable voltage and frequency conditions that could degrade instrwentation so that it would not perform the intended safety function.3/4.2.A Primary Containment and Reactor Building Isolation: Instrwentation that initiates primary contairvrent isolation.(Instrwents in Table 3.2.A and associated SRs in Table 4.2.A)(Exceptions listed below>3.3.6.1 3.3.6.2 3.3.7.1 3/4.3.2 3.3.6.1 3.3.6.2 3.3.7.1 Yes-3, 4 All Functions retained (with exceptions listed below)because the Functions actuate to mitigate the consequences of a DBA LOCA, Fuel Handling Accident or are considered risk significant and are retained in accordance with the NRC Final Policy Statement on Technical Specification Improvements. The isolation signals generated by the reactor building isolation instrwentation are implicitly assumed in the safety analyses to initiate closure of valves to limit offsite doses.3/4.2.A 3/4.2.A SGTS flow functions Reactor Building Isolation Timer Functions Relocated Hone Relocated Hone Hone None Ho See Secondary Contairment Isolation Instrwentation Justification for Change (LA3 for ISTS 3.3.6.2)See Secondary Contalwent Isolation Instrwentation Justification for Change (LA3 for ISTS 3.3.6.2)

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUHBER 3/4.2.B TITLE Core and Containment Cooling Systems-Initiation 5 Control: Instrunentation that initiates or controls the core and containment cooling systems (LPCI, CS, ADS, NPCI, and RCIC).(Instrunents in Table 3.2.B and associated SRs in Table 4.2.B)(Exceptions listed below)BFN ISTS NUMBER 3.3.5'I 3.3.5'3.3.6.1 STS REV.4 NUHBER 3/4.3.3 3/4.3.5 NUREG 1433 NUHBER 3.3.5.1 3.3.5.2 3.3.6.1 RETAINED/CR I TER ION FOR INCLUSION Yes-3, 4 BASIS FOR INCLUSION/EXCLUSION Functions retained (with exceptions listed below)because the various Functions actuate to mitigate the consequences of 8 DBA LOCA or are considered risk significant and are retained in accordance with the NRC Final Policy Statement on Technical Specification Improvements. Table 3/4.2.B Table 3/4.2.B Table 3/4.2.B Table 3/4.2.B Table 3/4.2.B Table 3/4.2.8 Drywell High Pressure (1<p<2~5 psig)Core Spray Sparger to Reactor Pressure Vessel d/p Trip System Bus Power Honitors: RHR (LPCI), CS, ADS, HPCI, and RCIC Trip Systems CS and RHR Discharge Pressure End of Cycle Recirculation Pump Trip: Instrunentation that trips the reactor recirculation pump to limit the consequences of a failure to scram (ATNS-RPT)~(Instruments in Table 3.2.8 and associated SRs in Table 4.2.8))CS and RHR Area Cooler Fan Thermostat Relocated None Relocated None Relocated Hone 3.3.4.1 3/4.3.4.1 Relocated Hone Relocated None None Hone Hone None 3.3.4.1 Hone No No No Yes-3 Ho See ECCS Instrunentation Justification for Change (R2 for ISIS 3.3.5.1)See Appendix A, Page A-3.See Appendix A, Page A-1 See ECCS Instrunentation Justification for Change (LA3 for ISTS 3.3.5.1)EOC-RPT aids the reactor scram in protecting fuel cladding integrity by ensuring the fuel cladding integrity Safety Limit is not exceeded during a load rejection or turbine trip transient. See ECCS Instrumentation Justification for Change (LA3 for ISTS 3.3.5.1)3/4.2.C Control Rod Block Actuation: Instrunentation that Initiates Control Rod Blocks.(Instrunents in Table 3.2.C and associated SRs in Table 4.2.C)(Exceptions listed below)3.3.2.1 3/4.3.6 3.3.2.1 Yes Control Rod Block Actuation Instrunentation functions to prevent violation of the HCPR Safety Limit and cladding plastic strain design limit during a single control rod withdrawal error event, ensures the initial conditions of the control rod drop accident analysis are not violated, and prevents inadvertent criticality when the reactor is shutdown (thereby preserving the safety analysis assumptions). SUHMARY DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER TITLE BFN ISTS NUMBER STS REV.4 NUMBER NUREG 1433 NUMBER RETAINED/CRITERION FOR IHCLUSIOH BASIS FOR INCLUSION/EXCLUSION Table 3/4.2.C Table 3/4.2.C Table 3/4.2.C Table 3/4.2.C 3/4.2.D APRM (Upscale Flow Bias, Upscale Startup Mode, APRM Downscale, and APRM Inoperative) IRM (Upscale, Downscale, Detector Not in Startup Position, Inoperative) SRM (Upscale, Downscale, Detector Not in Startup Position, Inoperative) Scram Discharge Votune High Level (Deleted)Relocated 3/4.3.6 Relocated 3/4.3.6 Relocated 3/4'.6 Relocated 3/4.3.6 Hone Hone None Hone No Ho No See Appendix A, Page A-4.See Appendix A, Page A-6.See Appendix A, Page A-7.See Appendix A, Page A-B.3/4.2.E 3/4.2.F 3/4'.M Drywell Leak Detection: Instrunentation that monitors drywell leakage Surveillance Instrunentation (Post Accident Monitoring Instrunents): Instrunentation that provide surveillance information.(Instrunents in Table 3.2.F and associated SRs in Table 4'.F)3.4.5 3.3.3.1 3/4.4.3.1 3.4 6 3/4 3.7.5 3 3 3 1 Yes-1 Yes-3 Leak detection instrunentation is used to indicate an abnormal condition of the reactor coolant pressure boundary.Regulatory Guide 1.97 Type A and Category 1 instruoents retained.See Appendix A.Page A-10, for full discussion of all instrunents in Table 3.2.F.3/4.2.G Control Room Isolation: Instrunentation that isolates the control room and initiates CREVs.(Instrm.nts in Table 3.2.G and associated SRs in Table 4.2.G)3.3.7.1 3/4.3.7 3.3.7.1 Yes-3 Functions actuate to maintain control room habitability so that operation can continue from the control room following a DBA.3/4.2.H 3/4.2.I 3/4.2.J Flood Protection Instrwentation Meteorological Monitoring Instrunentatton Seismic Monitoring Instrwentation Relocated Hone Retocated Hone Relocated None Hone None Hone Ho No No See Appendix A, page A-16.See Appendix A, page A-19.See Appendix A, page A-18.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUHBER 3/4'.K 3/4.2.L TITLE Explosive Gas Monitoring Instrunentation ATWS Recirculation Pump Trip BFN ISTS NUHBER Relocated 3.3.4.2 STS REV.4 NUMBER None 3/4.3.4.1 NUREG 1433 NUHBER None 3.3.4.2 RETAINED/CRITERION FOR INCLUSION Yes-4 BASIS FOR INCLUSION/EXCLUSION See Appendix A, page A-21.Part of the program required by BFN Specification 5.5.8.ATWS-RPT is being retained in accordance with NRC Final Policy Statement on I'echnical Specification Improvements due to risk significance. 3/4.3.A.1 3/4.3.A.2.a 3/4.3.A.2.b 3/4.3.A.2.c 3.3.A.2.d 3/4'.B.1 3.3.A.2.e/4.3.A.2 d Reactivity Hargin-Core Loading Reactivity Hargin-Inoperable Control Rods Scram Accunuiators 3.1.1 3.1.3 3.1.5 3/4.1.1 3/4.1.3'3/4.1.3.5 3/4.1.3.6 3/4.1.3.5 3/4.1.3.7 3.1.1 3.1.3 3.1.5 Yes-2 Yes-3 Yes-3 Shutdown Margin (SDH)is assumed as an initial condition for the control rod removal error during a refueling event and the fuel assembly insertion error during a refueling event.Control rods are part of the primary success path for mitigating the consequences of DBAs and transients. Same as above.3/4.3.B.2 Control Rod Housing Support Relocated 3/4.1.3.8 None No See Control Rod Operability Justification for Change (R1 for ISTS 3.1.3)3/4.3.8.3.b Rod Worth Minimizer 3.3.2.1 3/4.1.4.1 3.3.2.1 Yes-3 The RWH enforces the Banked Position Withdrawal Sequence (BPWS)to ensure that the initial conditions of the LOCA analysis are not violated.3/4.3.B.4 Minimum Count Rate for Control Rod Withdrawal 3.3.1.2 3/4.3.7.6 3.3.1.2 Yes Does not satisfy selection criteria, however is being retained because it is considered necessary for flux monitoring during shutdown, startup and refueling operations. 3/4.3.C 3/4.3.D Scram Insertion Times Reactivity Anomalies 3.1.3 3.1.4 3.1.2 3/4.1.3.2 3/4.1.3.3 3/4.1.3.4 3/4.1'3.1.3 3.1.4 3.1.2 Yes-3 Yes-2 Control rods are part of the primary success path for mitigating the consequences of DBAs and transients. The LOCA and transient analyses assune that control rods scram at a specified insertion rate.Not a measured process variable, but is important parameter that is used to confirm the acceptability of the accident analysis 3/4.3.F Scram Discharge Volune 3.1.8 3/4.1.3.1 3.1.8 Yes-3 The capability to insert the control rods ensures the assunptions used for the scram reactivity in the LOCA and transient analyses are maintained. The Scram Discharge Volune (SDV vent and drain valves contribute to the operability of the control rod scram function.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 3/4.4 3/4.5.A 3/4.5.8 3/4.5.B.11 3/4.5.B.12 3/4.5.8.13 3/4.5.C.1 3/4.5.C.2 3/4'.C.3 3/4.5.C.4 3/4.5.C.5 TITLE Standby Liquid Control System-CORE.AHD'CONTAINMEHT COOLING'.:," ,~SYSTEMS Core Spray System Residual Heat Removal System (RHRS)(LPCI AND Contaiment Cooling)RHR cross-connect capability between units RHR Service Mater and Emergency Equipment Cooling Hater Systems Standby Coolant Supply Capability BFN ISTS NUMBER 3.1.7 3.5.1 3.5.1 3.5.2 3.6.2.3 3.6.2.4 3.6.2.5 None 3.7.1 3.7.2 Hone STS REV.4 NUMBER 3/4.1.5 3/4.5.1 3/4.5.'I 3/4.5.2 3/4.6.2.2 3/4.6.2.3 Hone 3/4.7.1~1 3/4.7.1.2 3/4.7.1.3 None NUREG 1433 NUMBER 3.1.7 3.5.1 3.5.1 3.5.2 3.6.2.3 3.6.2.4 3.6.2.5 None 3.7.1 3.7.2 Hone RETAIHED/CRITERION FOR INCI.US ION Yes-4 Yes-3 Yes-3 Relocated Yes-3 Relocated BASIS FOR INCLUSION/EXCLUSION The Standby Liquid Control (SLC)is a backup system to the control rod scram function.This system is being retained per the NRC Final Policy Statement on Technical Specification Improvements due to the risk significance. Core Spray subsystems are part of the ECCS and function to provide cooling water to the reactor core to mitigate large Loss of Coolant Accidents. RHR Low Pressure Coolant Injection subsystems are part of the ECCS and function to provide cooling water to the reactor core to mitigate large Loss of Coolant Accidents. RHR Containment Cooling systems provide a reliable source of cooling water and functions to provide cooling to the primary contairment under post accident conditions. See Justification for Change Rl for BFH ISTS 3.5.1, ECCS.Designed for heat removal from various safety related systems following a DBA.As such, acts to mitigate the consequences of an accident.See Justification for Change R1 for BFH ISIS 3.7.1, RHRSH.3/4.5D Equipment Area Coolers Relocated Hone None No Relocated to the Bases as they are part of ECCS Operability. See ECCS Justification for Changes (3.5.1, LA4)3/4.5E 3/4.5F High Pressure Coolant Injection System Reactor Core Isolation Cooling System 3.5.1 3.5.3 3/4.5.1 3/4.7.4 3.5.1 3.5.3 Yes-3 Yes-4 The HPCI System is part of the ECCS and functions to mitigate small break Loss of Coolant Accidents. System retained in accordance with the NRC Final Policy Statement on Technical Specification improvements due to risk significance.

SUMMARY

DISPOSITION HATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUHBER 3/4~5G 3/4.5H 3/4.51 3/4.5J 3/4.5K 3/4.5L 3/4.5H TITLE Automatic Depressurization System (ADS)Haintenance of Filled Discharge Pipe Average Planar Linear Heat Generation Rate (HAPLHGR)Linear Heat Generation Rate (LHGR)Hinimm Critical Power Ratio (HCPR)APRH Setpoints Core Thermal-Hydraulic Stability BFN ISTS NUHBER 3.5.1 3.5.1 3.5.2 3.5.3 3.2.1 3.2.3 3.2.2 3.2.4 3.4.1 STS REV.4 NUHBER 3/4.5.1 3/4.5.1 3/4.5.2 3/4.5.4 3/4.2.1 3/4.2.4 3/4.2.3 3/4.2.2 3/4.4.1.1 3/4.4.1.3 NUREG 1433 NUHBER 3.5.1 3~5.1 3'.2 3.5.3 3.2.1 3.2.3 3.2'3.2.4 3.4.1 RETAINED/CRITERION FOR INCLUSION Yes-3 Yes-3, 4 Yes-2 Yes-2 Yes-2 Yes-2, 3 Yes-2 BASIS FOR INCLUSION/EXCLUSION The ADS is part of the ECCS and is designed to mitigate a small or mediun break Loss of Coolant Accident.The ADS acts to rapidly reduce reactor vessel pressure in a LOCA situation in which the HPCI System fails to automaticaily maintain reactor vessel water level.This depressurization enables the low-pressure emergency core cooling systems to deliver cooling water to the reactor core.This Specification ensures the operability of the ECCS and RCIC System, which function to mitigate the consequences of a LOCA (ECCS)or is required to be retained by the NRC Final Policy Statement on Technical Specification Isprovements (RCIC).The APLHGR limit is an initial condition in the safety analyses.The LHGR limit is an initial condition in the safety analyses.The HCPR limit is an initial condition in the safety analyses.The Operability of the APRHs and their setpoints is an initial condition of all safety analyses that assune rod insertion upon reactor scram.Recirculation loop flow is an initial condition in the safety analysis.3/4.6.A.1 3/4.6.A.2 3/4.6.A.3 3/4.6.A.4 3/3.6.A.S 3/4.6.A.6 3/4.6.A.7 Thermal and Pressurization Limitations Idle Recirculation Loop Startup 3'.9 3.4.9 3/4.4.6.1 3.4.'IO 3/4.4.6.1 3.4.10 Yes-2 Yes-2 Establishes initial conditions such that operation is prohibited in areas or at temperature rate changes that might cause undetected flaws to propagate in turn challenging the reactor coolant pressure boundary integrity. Same as above.3/4.6.8.1 3/4.6.8.2 3/4.6.8.3 3/4.6.8.4 3/4.6.8.5 Coolant Chemistry Relocated 3/4.4.4 Hone No See Appendix A, Page A-12.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 3/4.6.B.6 3/4.6.C.1 3/4.6.C.2 3/4.6.D 3/4.6E 3/4.6F 3/4.6G 3/4'H TITLE Specific Activity Coolant Leakage Leakage Detection Systems Relief Valves Jet Pumps Recirculation Pump Operation Structural Integrity Snubbers BFN ISTS NUMBER 3.4.6 3.4.4 3.4.5 3.4.3 3.4.2 3.4.1 Relocated Relocated STS REV.4 NUHBER 3/4.4.5 3/4.4.3.1 3/4.4.3.2 3/4.4.2.1 3/4.4.1.2 3/4.4.1.1 3/4.4.1.3 3/4.4.8 3/4.7.5 NUREG 1433 NUMBER 3.4.7 3.4.4 3.4.6 3.4.3 3.4.2 3.4.1 None None RETAINED/CRITERION FOR IHCLUSION Yes-2 Yes-1,2 Yes-1, 2 Yes-3 Yes-2 Yes-2 Ho BASIS FOR INCLUSION/EXCLUSION The specific activity in the reactor coolant is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB)outside contaiwent. Leakage beyond limits would indicate an abnormal condition of the reactor coolant pressure boundary.Operation in this condition may result in reactor coolant pressure boundary failure.Leakage detection instrunents are used to indicate an abnormal condition of the reactor coolant pressure boundary.Same as above.The Safety and Relief Valves are assuned to operate to maintain the reactor pressure below design limits.Jet Purp operability is explicitly assed in the design basis LOCA to assure adequate core reflood capability. Recirculation loop flow is an initial condition in the safety analysis.See Appendix A, Page A-13 See Justification for Change (CTS 3.6.H/4.6.H, LA1)for relocating snubbers in CTS 3.6.H/4.6.H.(Spec 3.4 markup)3/4.7.A.1 3/4.7.A.2 3/4.7.A.3 3/4.7.A.4 Suppression Chamber Primary Contairvaent Integrity Pressure Suppression Chamber-Reactor Building Vacuun Breakers Drywell-pressure Suppression Chamber Vacua Breakers 3.6.2.1 3.6.2.2 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.5 3.6.1.6 3/4.6.2.1 3/4.6.1.1 3/4.6.1'3/4.6.1.3 3/4.6.1'3/4.6.4.2 3/4.6.4.1 3.6.2'3.6.2.2 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.7 3.6.1.8 Yes-2, 3 Yes-3 Yes-3 Yes-3 The suppression pool water voiune and terrperature are initial conditions in the DBA LOCA contairvaent response analysis and mitigate the consequences of a DBA.Primary contairvaent functions to mitigate the consequences of a DBA.Primary contaiment leakage is an assumption utilized in the LOCA safety analysis to ensure primary contairment operability. Pressure suppression chamber to reactor building vacrxmr breaker operation is relied upon to limit a negative pressure differential, secondary to primary contairvaent, that could challenge primary contairment integrity. Drywell-pressure suppression chamber vacrxmr breaker operation is assumed in the LOCA analysis to limit drywell pressure thereby ensuring primary contairment integrity.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 3/4.7.A.S Oxygen Concentration BFN ISTS NUMBER 3.6.3.2 STS REV.4 NUMBER 3/4.6.6.4 NUREG 1433 NUMBER 3.6.3.2 RETAINED/CR I TER ION FOR I N CLUB I ON Yes-2 BASIS FOR INCLUSION/EXCLUSIOH Oxygen concentration is limited such that, when combined with hydrogen (that is postulated to evolve following a LOCA), the total explosive gas concentration remains below explosive levels.Therefore, primary contairvnent integrity is maintained. 3/4.7.A.6 3/4.78 3/4.7C 3/4.7D 3.7.F.3.a Drywell-suppression Chamber Differential Pressure Standby Gas Treatment System Secondary Contaireent Primary Contairment Isolation Valves 3.6.2.6 3.6.4'3.6.4.1 3.6.4.2 3.6.1.3 3/4.6.2.4 3/4.6.5.3 3/4.6.5.1 3/4.6.5.2 3/4.6.3 3/4.6.1.8 3.6.2.5 3.6.4.3 3.6.4.1 3.6.4.2 3.6.1.3 Yes-2 Yes-3 Yes-3 Yes-3 Dryweii-suppression Chamber Differential Pressure is an initial condition in the DBA LOCA contairment response analysis.System functions following a DBA to limit offsite releases.Secondary contaiment integrity is relied on to limit the offsite dose during an accident by ensuring a release to contairment is delayed and treated prior to release to the enviroreent. Damper operation within time limits establishes secondary contaim.nt and limits offsite releases to acceptable values.Isolation valves function to limit DBA consequences. 3/4.7F Primary Contairment Purge System Relocated Hone None No See Justification for Change (CTS 3.7.F/4.7.F, R1 at the end of markup for proposed BFH ISTS 3.6)for relocating primary contairment purge system.3/4.7E 3/4.7.G Control Room Emergency Ventilation Contairment Atmosphere Dilution System (CAD)3.7.3 3.6.3~1 3/4.7.2 3/4.6.6.2 3.7.4 3.6.3.4 Yes-3 Yes-3 Maintains habitability of the control room so that operators can remain in the control room following an-accident. As such, it mitigates the consequences of an accident by allowing the operators to continue accident mitigat'Ion activities from the control room.System ensures oxygen concentration is maintained below the explosive level following a LOCA by inerting the dryweli with nitrogen.Therefore, contairment integrity is maintained. 3.8.A.S 3/4.8.A.6 Liquid Holdup Tanks 5.5.8 Hone 5.5.8 Yes Although this Specification does not meet any criteria of the HRC Final Policy Statement, it has been retained in accordance with NRC Letter from II.T.Russell to the Industry ITS chairpersons, dated October 25, 1993.

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 3.8.8.9 3.8.B.10 4.8.8.5 3/4.8.E TITLE Airborne Effluents-Explosive Gas Mixture Miscellaneous Radioactive Materials Sources BFH ISTS NUMBER 5.5.8 Relocated STS REV.4 NUMBER None 3/4.7.6 HUREG 1433 NUMBER 5.5.8 None RETAINED/CRITERION FOR I NCLUS IOH Yes No BASIS FOR INCLUSION/EXCLUSION See Appendix A, page A-20.This is a requirement of the program required by BFN Specification 5.5.8.See Appendix A, page A-'I7.3/4.9.A.1 3.9.A.2 3.9.A.6 3/4.9.8.1 3/4.9.8.3 3.9.8.15 3.9.A.3/4.9.A.4.D 4.9.A.S 3/4.9.8.2 3/4.9.B.4 3/4.9.8.5 3/4.9.B.6 3.9.8.12 3.9.8.13 3.9.B.14 3.9.B.15 3.9.A.4/4.9.A.2 3.9.8.7 3.9.8.8 3.9.B.15 3.9.A.5/4.9.A.3.a 3/4.9.C.1 3.9.C.2 3.9.C.3 3.9.C.4 3/4.9.D Auxiliary Electrical Equipment-A.C.Sources Operating Auxiliary Electrical Equipment-Buses and Boards Available D.C.Power System Logic Systems A.C.Sources-Operation in Cold Shutdown Onsite Electrical Power Distribution -Shutdown Unit 3 Diesel Generators Required for Unit 2 Operation 3.8.1 3.8.7 3.8.4 3.8.1 3.3.5.1 3.8.2 3.8.8 3.8.1 3.8.2 3/4.8.1.1 3/4.8.3.1 3/4.8.3.1 3/4.8.1.1 3/4.8.1.2 3/4.8.1.2 3/4.8.1.1 3/4.8.1.2 3.8.1 3.8.9 3.8.4 3.8.1 3.3.5.1 3.8.2 3.8.10 3.8.1 3.8.2 Yes-3 Yes-3 Yes-3 Yes-3 Yes-3 Yes-3 Yes-3 The operability of the AC power sources is part of the primary success path of the accident analyses.The operability of the distribution system is part of the primary success path of the accident analyses.The operability of the DC subsystems is consistent with the initial assumptions of the accident analyses.Required to mitigate the consequences of a DBA.Same as above.Same as above.Same as above.10

SUMMARY

DISPOSITION MATRIX FOR BFN UNITS I, 2, AND 3 CURRENT TS NUMBER 3/4.10.A.1 3/4.10.A.2 TITLE Refueling Operations-Interlocks BFH ISTS NUMBER 3.9.1 3.9.2 3.9.3 STS REV.4 NUMBER 3/4.9.'I NUREG 1433 NUHBER 3.9.1 3.9.2 3.9.3 RETAINED/CR I TER IOH FOR INCLUSION Yes-3 BASIS FOR INCLUSION/EXCLUSION The refueling interlocks protect against prompt reactivity excursions during the Refuel Hode.The safety analyses for the control rod removal error during refueling end the fuel assembly insertion error during refueling assme the functioning of the refueling interlocks. 3.10.A.3 3.'IO.A.4 Refueling Platform Equipment Interlocks Relocated 3/4.9.7 None No See Appendix A, Page A-22.3/4.10.A.5 3/4.10.A.6 3/4.10.A.7 3/4.10.B Refueling Operations -Single Control Rod Naintenance Refueling Operations -Removal of Two Control Rods Refueling Operations -Removal of Any Number of Control Rods Refueling Operations -Core Monitoring 3.10.4 3.10.5 3.10.6 3.3.1.2 3/4.9.10~2 3/4.9.10.2 3/4.9.10'3/4.9.2 3.10.4 3.10.4 3.10.6 3.3.1.2 Yes Yes Yes Yes This requirement is being retained to allow relaxation of certain Limiting Conditions for operation (LCOs)under specific conditions to allow testing and maintenance. This requirement is directly related to several LCOS.Direct application of the Technical Specification selection criteria is not appropriate. However, this requirement, directly tied to LCOs that remain in Technical Specifications, will also remain in Technical Specifications. Same as above.Same as above.Does not satisfy criteria for inclusion but is retained because it is considered necessary for flux monitoring during shutdown, startup, and refueling operations. 3/4.10.D 3/4.10.E 3.10'Refueling Operations -Reactor Building Crane Refueling Operations -Spent Fuel Cask Relocated 3/4.9.6 Relocated 3/4.9.6 None None No See Appendix A, Page A-14 See Appendix A, Page A-15 5.0 Major Design Features 4.0 5.0 4.0 Yes Application of Technical Specification selection criteria is not appropriate. However, Design Features will be included in Technical Specifications as required by 10 CFR 50.36a.6.0 Administrative Controls 5.0 6.0 5.0 Yes Application of Technical Specification selection criteria is not appropriate. However, Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36a.11 APPENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION APPENDIX A TABLE 3/4.2.B TRIP SYSTEM BUS POWER MONITORS FOR THE RHR (LPCI), CORE SPRAY, ADS, HPCI AND RCIC TRIP SYSTEMS LCO Statement: The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.Table 3.2.B Instrumentation that Initiates or Controls the Core and Containment Coolin S stems 3/4.2-17 3/4.2-17 3/4.2-17 3/4.2-18 3/4.2-18 Discussion: RHR (LPCI)Trip System Bus Power Monitor Core Spray Trip System Bus Power Monitor-ADS Trip System Bus Power Monitor HPCI Trip System Bus Power Monitor RCIC Trip System Bus Power Monitor The Trip System Bus Power Monitors for the RHR (LPCI), Core Spray, ADS, HPCI and RCIC trip systems alarm if a fault is detected in the power system to the appropriate systems logic.No design basis accident (DBA)or transient analyses takes credit for the Trip System Bus Power Monitors.This instrumentation provides a monitoring/alarm function only.Com arison to Screenin Criteria: 2.3.The Trip System Bus Power Monitors are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.The Trip System Bus Power Monitors are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.The Trip System Bus Power Monitors are not part of the primary success path that 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.As discussed in Sections 3.5 and 6 of NEDO-31466 and summarized in Table 4-1 (item 106)of NED0-31466, Supplement 1, and verified by TVA, the loss of the RHR (LPCI), Core Spray, ADS, HPCI and RCIC Trip System Bus Power Monitors was found to be a non-significant risk contributor to core damage frequency and offsite releases.A-1 APPENDIX A TABLE 3/4.2.B TRIP SYSTEM BUS POWER MONITORS FOR THE RHR (LPCI), CORE SPRAY, ADS, HPCI AND RCIC TRIP SYSTEMS (cont'd.)Conclusion: Since the screening criteria have not been satisfied, the RHR (LPCI), Core Spray, ADS, HPCI and RCIC Trip System Bus Power Monitors LCO and Surveillances may be relocated to a licensee controlled document.A-2 APPENDIX A TABLE 3/4.2.8 CORE SPRAY SPARGER TO REACTOR PRESSURE VESSEL d/p LCO Statement: The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.Table 3.2.B 3/4.2-17 Discussion: Instrumentation that Initiates or Controls the Core and Containment Coolin S stems Core Spray Sparger to Reactor Pressure Vessel d/p This instrumentation measures the differential pressure between the core spray sparger and the reactor pressure vessel above the core plate and alarms if a break is detected.This Function does not actuate any equipment; it provides an alarm function only.This Function monitors the integrity of the core spray system piping in the reactor annulus region which would not otherwise be apparent to the operators. It is not credited in the accident analysis.Com arison to Screenin Criteria: 2.This instrumentation is not the primary method for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.This instrumentation is not a process variable 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.This instrumentation is not part of the primary success path that 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.As discussed in Appendix B, Page 1, TVA found the loss of the Core Spray Sparger to Reactor Pressure Vessel d/p Instrumentation to be a non-significant risk contributor to core damage frequency and offsite releases.r

Conclusion:

Since the screening criteria have not been satisfied, the Core Spray Sparger to Reactor Pressure Vessel d/p Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-3 APPENDIX A 2.1.A.l.c APRH ROD BLOCK TRIP SETTING TABLE 3/4.2.C CONTROL ROD BLOCKS-APRH UPSCALE (FLOW BIASED, STARTUP MODE), APRH DOWNSCALE LSSS Statement: The limiting safety system settings shall be as specified below: A.l.c The APRH Rod Block Trip Setting shall be less than or equal to the limit specified in the COLR LCO Statement: The limiting conditions of operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.Table 3.2.C Instrumentation that Initiates or Controls the Core and Containment Coolin S stems 3/4.2-25 3/4.2-25 3/4.2-25 3/4.2-25 Discussion: APRH Upscale (Flow Biased)APRM Upscale (Startup Mode)APRM Downscale APRM Inoperative The Average Power Range Monitor (APRM)control rod blocks function to prevent a control rod withdrawal error during power range operations using LPRH signals to create the APRH rod block signal.APRHs provide information about the average core power and APRH rod blocks are not assumed to mitigate a DBA or transient. Com arison to Screenin Criteria: 2.3.The APRH control rod blocks are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.The APRH control rod block instrumentation is not used to monitor a process variable 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.The APRH control rod blocks are not part of the primary success path that 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.A-4 2.1.A.l.c TABLE 3/4.2.C APPENDIX A APRH ROD BLOCK TRIP SETTING CONTROL ROD BLOCKS-APR.UPSCALE (FLOW BIASED, STARTUP MODE), APRH DOWNSCALE (cont'd.)As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 135)of NEDO 31466, and verified by TVA, the loss of the APRH control rod block functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the APRH Control Rod Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-5 APPENDIX A TABLE 3/4.2.C CONTROL ROD BLOCKS-IRH UPSCALE, IRH DOWNSCALE, IRH DETECTOR NOT IN STARTUP POSITION, IRH INOPERATIVE LCO Statement: The limiting conditions of operation for the instrumentation that initiate control rod blocks are given in Table 3.2.C.Table 3.2.C Instrumentation that Initiates or Controls the Core and Containment Coolin S stems 3/4.2-25 3/4.2-25 3/4.2-25 3/4.2-25 Discussion: IRH Upscale IRH Downscale IRH Detector Not In Startup Position IRH Inoperative The Intermediate Range Monitor (IRH)control rod blocks function to prevent a control rod withdrawal error during reactor startup using IRH signals to create the rod block signal.IRHs are provided to monitor the neutron flux levels during refueling and startup conditions. No design basis accident or transient analysis takes credit for rod block signals initiated by IRHs.Com arison to Screenin Criteria: 1.The IRH control rod blocks are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2.3.The IRH control rod block instrumentation is not a process variable 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.The IRH control rod blocks are not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 138)of NED0-31466, and verified by TVA, the loss of the IRH control rod block functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the IRH Control Rod t Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-6 APPENDIX A TABLE 3/4.2.C CONTROL ROD BLOCKS-SRM UPSCALE, SRM DOWNSCALE, SRH DETECTOR NOT IN STARTUP POSITION, SRM INOPERATIVE LCO Statement: The limiting conditions of operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.Table 3.2.C Instrumentation that Initiates or Controls the Core and Containment Coolin S stems 3/4.2-25 3/4.2-25 3/4.2-25 3/4.2-25 Discussion: SRH Upscale SRM Downscale SRM Detector Not In Startup Position SRM Inoperative The Source Range Monitor (SRM)control rod blocks function to prevent a control rod withdrawal error during reactor startup using SRM signals to create the rod block signal.SRH signals are used to monitor the neutron flux levels during refueling, shutdown, and startup conditions. No design basis accident or transient analysis takes credit for rod block signals initiated by the SRMs.Com arison to Screenin Criteria: 1.The SRM control rod blocks are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2.3.The SRH control rod block instrumentation is not a process variable 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.The SRM control rod blocks are not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 137)of NED0-31466, and verified by TVA, the loss of the SRM control rod block functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the SRH Control Rod t Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-7 APPENDIX A TABLE 3.2.C CONTROL ROD BLOCKS-SCRAM DISCHARGE INSTRUMENT VOLUME HIGH LEVEL LCO Statement: The limiting conditions for operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.Table 3.2.C Instrumentation that Initiate Control Rod Blocks 3/4.2-25 Discussion: Scram Discharge Instrument Volume High Level The Scram Discharge Volume (SDV)control rod block functions to prevent control rod withdrawals during power range operations, utilizing SDV high level signals to create the rod block signal, if water is accumulating in the SDV.The purpose of monitoring the SDV water level is to ensure that there is sufficient volume remaining to contain the water discharged by the control rod drive during a scram, thus ensuring that the control rods will be able to insert fully.This rod block signal provides an indication to the operator that water is accumulating in the SDV and prevents further control rod withdrawals. With continued water accumulation, a reactor protection system initiated scram signal will occur.Thus, the SDV water level rod block signal provides an opportunity for the operator to take action to avoid a subsequent scram.No design basis accident (DBA)or transient analysis takes credit.for rod block signals initiated by the SDV high level instrumentation. Com arison to Screenin Criteria: 2.3.4.The SDV control rod block is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.The SDV control rod block instrumentation is not a process variable 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.The SDV control rod block is not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 139)of NED0-31466, and verified by TVA, the loss of the Scram Discharge Volume High Level Control Rod Block Instrumentation was found to be a nonsignificant risk contributor to core damage frequency and offsite releases.A-8 APPENDIX A TABLE 3.2.C CONTROL ROD BLOCKS-SCRAM DISCHARGE INSTRUMENT VOLUME HIGH LEVEL (cont'd.)Conclusion: Since the screening criteria have not been satisfied, the Scram Discharge Instrument Volume High Level Control Rod Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-9 APPENDIX A 3/4.2.F SURVEILLANCE INSTRUHENTATION LCO Statement: The limiting conditions for the instrumentation that provides surveillance information readouts are given in Table 3.2.F Table 3.2.F Surveillance Instrumentation 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-31 3/4.2-32 3/4.2-32 3/4.2-32 3/4.2-32 3/4.2-32 3/4.2-32 3/4.2-32 3/4.2-32 Discussion: Reactor Water Level Reactor Pressure Drywell Pressure Drywell Air Temperature Suppression Chamber Air Temperature Control Rod Position Neutron Honitoring Drywell Pressure Alarm Drywell Temperature and Pressure and Timer CAD Tank Level Drywell and Torus Hydrogen Concentration Drywell to Suppression Chamber Differential Pressure Relief Valve Tailpipe Thermocouple Temperature or Acoustic Honitor on Relief Valve Tailpipe Primary Containment High Range Radiation Honitors Drywell Pressure-Wide Range Suppression Chamber Water Level-Wide Range Suppression Pool Bulk Temperature Wide Range Gaseous Effluent Radiation Honitor Each individual accident monitoring parameter has a specific purpose, however, the general purpose for all accident monitoring instrumentation is to provide sufficient information to confirm an accident is proceeding per prediction, i.e., automatic safety systems are performing properly, and deviations from expected accident course are minimal.Com arison to Screenin Criteria: The NRC position on application of the deterministic screening criteria to post-accident monitoring instrumentation is documented in letter dated Hay 7, 1988 from T.E.Hurley (NRC)to R.F.Janecek (BWROG).The position taken was that the post-accident monitoring instrumentation table list should contain, on a plant specific basis, all Regulatory Guide 1.97 Type A instruments specified in the plants SER on Regulatory Guide 1.97, and all Regulatory Guide 1.97 Category 1 instruments. Accordingly, this position has been applied to the BFN Regulatory Guide 1.97 instruments. Those instruments not meeting this criteria have been relocated from the Technical Specifications to a licensee controlled document. APPENDIX A 3/3.2.F SURVEILLANCE INSTRUMENTATION (cont'd.)The following summarizes the BFN position for those instruments currently in Technical Specifications. From NRC SER dated April 30, 1984,

Subject:

Conformance to RG 1.97 Cate or 1 or T e A Variables 1.2.3.5.6.7.8.Reactor Pressure Reactor Vessel Water Level (wide range, accident range)Suppression Pool Water Temperature Suppression Pool Water Level (wide range)Drywell Pressure (normal range, wide range)Drywell Air Temperature Primary Containment Area Radiation Drywell and Torus Hydrogen Concentration For other post-accident monitoring instrumentation currently in Technical Specifications, their loss is not considered risk significant since the variable they monitored did not qualify as a Type A (one that is important to safety and needed by the operator, so that the operator can perform necessary manual actions)or Category 1 variable.Conclusion: Since the screening criteria have not been satisfied for instruments that do not meet Regulatory Guide 1.97 Type A variable requirements or Category 1 variable Type A instruments, their associated LCO and Surveillances will be relocated to a licensee controlled document.The instruments to be relocated are as follows: l.2.3.4, 5.6.7.8.9.Drywell Temperature and Pressure Timer Suppression Chamber Air Temperature Control Rod Position Neutron Monitoring Drywell Pressure Alarm CAD Tank Level Drywell to Suppression Chamber Differential Pressure Relief Valve Tailpipe Thermocouple Temperature or Acoustic Monitor on Relief Valve Tailpipe Wide Range Gaseous Effluent Radiation Monitor APPENDIX A 3/4.6.B.1-5 PRIMARY SYSTEM BOUNDARY-COOLANT CHEMISTRY LCO Statement: The following limits shall be observed for reactor water quality prior to any startup and when operating at rated pressure: a)Conductivity at 25'C-2.0 mho/cm b)Chloride concentration -0.1 ppm Discussion: Poor reactor coolant water chemistry contributes to the long-term degradation of system materials and, thus, is not of immediate importance to the plant operator.Reactor coolant water chemistry is maintained to reduce the possibility of failure in the reactor coolant system pressure boundary caused by corrosion. In summary, the chemistry monitoring activity is of a long term preventive purpose rather than mitigative. Com arison to Screenin Criteria: 1.Reactor coolant water chemistry is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2.Reactor coolant water chemistry is not a process variable 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.Reactor coolant water chemistry is not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 211)of NED0-31466, and verified by TVA, Coolant Chemistry requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the Coolant Chemistry (Conductivity and Chloride)LCO and Surveillances may be relocated to a licensee controlled document. APPENDIX A 3/4.6.G PRIMARY SYSTEM BOUNDARY-STRUCTURAL INTEGRITY LCO Statement: The structural integrity of the primary system boundary shall be maintained at the level required by the original acceptance standards throughout the life of the station.Discussion: The inspection programs for ASHE Code Class 1, 2, and 3 components ensure that the structural integrity of those components will be maintained throughout the components life.Operability of the primary system boundary is ensured by separate Technical Specifications and therefore, the inspections are not required to be retained in the Technical Specifications. This Technical Specification is more directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions. However, it is not necessary to retain this Specification to ensure the operability of the primary system boundary.Com arison to Screenin Criteria: 1.The inspections stipulated by this Specification are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2.3.The inspections stipulated by this Specification do not monitor process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.The ASNE Code Class 1, 2, and 3 components inspected per this Specification are assumed to function to mitigate a DBA.Their capability to perform this function is addressed by other Technical Specifications. This Technical Specification, however, only specifies inspection requirements for these components; and these inspections can only be performed when the plant is shutdown.Therefore, Criterion 3 is not satisfied. 4, As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 216)of NED0-31466, and verified by TVA, the lack of a Structural Integrity Specification was found to be a non-significant risk contributor to core damage frequency and offsite releases since the requirement is currently covered by 10 CFR 50.55a and the Inservice Inspection Program.Conclusion: Since the screening criteria have not been satisfied, the Structural Integrity LCO and Surveillance may be relocated to a licensee controlled document.A-13 APPENDIX A 3/4.10.D REFUELING OPERATIONS -REACTOR BUILDING CRANE LCO Statement The reactor building crane shall be OPERABLE: a.When a spent fuel cask is handled.b.Whenever new or spent fuel is handled with the 5-ton hoist.Discussion: The reactor building crane and 125 ton hoist are required to be operable for handling of the spent fuel in the reactor building.This LCO specifies minimum operability requirements to prevent damage to the refueling platform equipment and core internals. The crane is not assumed to function to mitigate the consequences of a DBA.Com arison to Screenin Criteria: 1.The reactor building crane is not used, nor is it capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).2.The reactor building crane is not a process variable that is an initial condition of a DBA or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.3.4.The reactor building crane is 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 287)of NED0-31466, and verified by TVA, Reactor building crane requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the LCO and associated surveillance may be relocated to the Technical Requirements Manual.A-14 APPENDIX A 3/4.10.E REFUELING OPERATIONS -SPENT FUEL CRANE 3.10.F LCO Statements Spent Fuel Cask Upon receipt, an empty fuel cask shall not be lifted until a visual inspection is made of the cask-lifting trunnions and fastening connection has been conducted. Spent Fuel Cask Handling-Refueling Floor Administrative control shall be exercised to limit the height the spent fuel cask is raised above the refueling floor by the reactor building crane to 6 inches, except for entry into the cask decontamination chamber where height above the floor will be approximately 3 feet.The spent fuel cask yoke safety links shall be properly positioned at all times except when the cask is in the decontamination chamber.Discussion: BFN analysis has been performed to address the handling of spent fuel cask.However, BFN currently does not have the need to handle spent fuel cask.Therefore, these LCOs serve no useful purpose and should be deleted.Com arison to Screenin Criteria: 2.3.The Spent fuel cask and spent fuel cask handling controls are not used to detect, and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).The Spent fuel cask and spent fuel cask handling controls are not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.The Spent fuel cask and spent fuel cask handling controls 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 287)of NED0-31466, and verified by TVA, Spent fuel cask and spent fuel cask handling control requirements not being met'was found to be a non-significant risk contributor to core damage frequency and offsite releases. APPENDIX A 3/4.10.E REFUELING OPERATIONS -SPENT FUEL CRANE 3.10.F (cont'd.)Conclusion: Since the screening criteria have not been satisfied and the LCOs serve no useful purpose, the LCOs and associated surveillance may be deleted.A-16 APPENDIX A 3/4.2.H FLOOD PROTECTION INSTRUMENTATION LCO Statement: The unit shall be shutdown and placed in the cold condition when Wheeler Reservoir lake stage rises to a level such that water from the reservoir begins to run across the pumping station deck at elevation 565.Requirements for the instrumentation that monitors the reservoir level are given in Table 3.2.H.Discussion: Provides capability to predict flood levels of large magnitudes which allows the plant to take advantage of advance warning to take appropriate action when reservoir levels above normal pool are predicted. Com arison to Screenin Criteria: 1.The Reservoir Level Monitoring instrumentation is not used to detect, and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).t 2.The Reservoir Level Monitoring instrumentation is not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.3.4.The Reservoir Level Monitoring instrumentation is 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 273)of NE00-31466, and verified by TVA, Reservoir Level Monitoring instrumentation requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the LCO and associated surveillance may be relocated to a licensee controlled document.A-17 APPENDIX A MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES LCO Statement: The leakage test shall be capable of detecting presence of 0.005 microcurie of radioactive material on the test sample.Discussion: The limitations on sealed source contamination are intended to ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation. This is done by imposing a limitation on the maximum amount of removable contamination on each sealed source.This requirement and the associated surveillance requirements bear no relation to the conditions or limitations which are necessary to ensure safe reactor operation. Com arison to Screenin Criteria: 3.4.Miscellaneous radioactive materials sources requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.Miscellaneous radioactive materials sources requirements are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Miscellaneous radioactive materials sources requirements are not part of the primary success path that function or actuate to mitigate a DBA or transient that either assumes the failure of'or presents a challenge to the integrity of a fission product barrier.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 267)of NEDO 31466, and verified by TVA, the Miscellaneous Radioactive Materials Sources requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the Miscellaneous Radioactive Materials Sources LCO and Surveillances may be relocated to a licensee controlled document.A-18 APPENDIX A 3/4.2.J SEISMIC MONITORING INSTRUMENTATION LCO Statement: The seismic monitoring instrumentation shown in Table 3.2.J shall be operable.Discussion: In the event of an earthquake, seismic monitoring instrumentation is required to determine the magnitude of the seismic event.These instruments do not perform any automatic action.They are used to measure the magnitude of the seismic event for comparison to the design basis of the plant to ensure the design margins for plant equipment and structures have not been violated.Since the determination of the magnitude of the seismic event is performed after the event has occurred, this instrumentation has no bearing on the mitigation of any design basis accident (DBA)or transient. Com arison to Screenin Criteria: 1.Seismic 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.Seismic monitoring instrumentation is not a process variable 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.Seismic monitoring instrumentation is not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 151)of NED0-31466, and verified by TVA, the loss of the Seismic Monitoring instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the Seismic Monitoring Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.A-19 3.2.F/4.2.F LCO Statement APPENDIX A METEOROLOGICAL MONITORING INSTRUHENTAT ION The meteorological monitoring instrumentation listed in Table 3.2.I shall be OPERABLE at all times.Discussion: Ensures that there is a sufficient amount of data available to estimate potential radiological doses.There are no automatic actions during any event that these instruments perform, nor do they actuate to mitigate a DBA or transient. Com arison to Screenin Criteria: 1.The Meteorological Monitoring instrumentation is not used to detect, and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).The Meteorological Monitoring instrumentation is 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.2.The Meteorological Monitoring instrumentation is not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses that either assumes the failure of or presents a t challenge to the integrity of a fission product barrier.3.4.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 151)of NEDO-31466 (Table 4-1 and 6-3, Item 152), the loss of meteorological monitoring instrumentation is a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the LCO and associated surveillance may be relocated to the Technical Requirements Manual.A-20 APPENDIX A 3.8.B.9, 10 4.8.B.5 RADIOACTIVE MATERIALS-AIRBORNE EFFLUENTS, EXPLOSIVE GAS MIXTURE LCO Statement: The concentration of hydrogen downstream of the recombiners shall be limited to less than or equal to 4%by volume.Discussion: The explosive gas mixture Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limit of hydrogen.However, the waste gas holdup system is designed to contain detonations and will not affect the function of any safety related equipment. The concentration of hydrogen in the offgas stream is not an initial assumption of any design basis accident (DBA)or transient analysis.Com arison to Screenin Criteria: 2.3.The explosive gas mixture requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.The explosive gas mixture requirements are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.The explosive gas mixture requirements are not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 306)of NED0-31466, and verified by TVA, an explosive gas mixture in the waste gas holdup system was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the Explosive Gas Mixture LCO and Surveillances may be relocated to a licensee controlled document.A-21 3.2.F/4.2.F LCO Statement APPENDIX A EXPLOSIVE GAS MONITORING INSTRUMENTATION The explosive gas monitoring instrumentation listed in Table 3.2.K shall be OPERABLE with the applicability as shown in Tables 3.2.K/4.2.K. Discussion: The explosive gas monitoring instrumentation is provided to ensure that the concentration of potentially explosive gas mixtures contained in the gaseous radwaste treatment system is adequately monitored, which will help ensure that the concentration is maintained below the flammability limit of hydrogen.However, the offgas system is designed to contain detonations and will not affect the function of safety related equipment. The concentration of hydrogen in the offgas system is not an initial assumption of any design basis accident or transient analysis.Com arison to Screenin Criteria: 3.The 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.The explosive gas monitoring instrumentation is not used to monitor a process variables that is an initial conditions of a DBA or transient. Excessive system effluent is not an indication of a DBA or transient. The explosive gas monitoring instrumentation is not part of the primary success path that 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (Item 306)of NED0-31466, and verified by TVA, an explosive gas mixture in the waste gas holdup system was found to be a non-significant risk contributor to core damage frequency and offsite releases.Conclusion: Since the screening criteria have not been satisfied, the Explosive Gas Mixture LCO and Surveillances may be relocated to a licensee controlled document.A-22 3.10.A.3&4 LCO Statement APPENDIX A REFUELING PLATFORM EQUIPMENT INTERLOCKS Refueling Interlocks 3.The fuel grapple hoist load switch shall be set at (1,000 lbs.If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service platform hoist is to be used for handling fuel with the head off the reactor vessel, the load limit switch on the hoist to be used shall be set at<400 lbs..Discussion: Specifies minimum operability requirements. Designed to provide the capabilities to prevent damage to the refueling platform equipment and core internals, they are not assumed to function to mitigate the consequences of a DBA.Com arison to Screenin Criteria: 1.Refueling platform equipment interlocks are not used to detect, and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).2.3.Refueling platform equipment interlocks are not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Refueling platform equipment interlocks 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.As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 306)of NED0-31466, and verified by TVA, the loss of the refueling platform equipment interlocks is a non-significant contributor to core damage frequency and offsite release.Conclusion: Since the screening criteria have not been satisfied, the LCO and associated surveillance may be relocated to a licensee controlled document.. A-23 APPENDIX B BFN SPECIFIC RISK SIGNIFICANT EVALUATIONS APPENDIX B TABLE 3/4.2.B CORE SPRAY SPARGER TO REACTOR PRESSURE VESSEL d/p LCO Statement: The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.Table 3.2.B Instrumentation that Initiates or Controls the Core and Containment Coolin S stems 3/4.2-17 Core Spray Sparger to Reactor Pressure Vessel d/p Descri tion of Re uirement: This instrumentation measures the differential pressure between the core spray sparger and the reactor pressure vessel above the core plate and alarms if a break is detected.This instrumentation does not actuate any equipment. Risk Justification: The function of the instrumentation is to identify a break in the core spray sparger.The probability of a pipe break (EPRI TR-100380) is extremely low, therefore the relative probability as defined in NEDO-31466 is low.LOCAs represent a small contribution to the BFN core damage frequency (CDF).A break in the sparger in the reactor pressure vessel would, without a LOCA, provide injection to the core.Given the success of injection with a break in non-LOCA accidents and the small contribution of LOCA to the CDF, the relative significance from an offsite radiological dose perspective would be low.The risk category would therefore be considered non-significant (NS).Relative Probabilit Rl 1 Rl 111~ltikC Low Low APPENDIX 8 3/4.2.LCO Statement FLOOD PROTECTION The unit shall be shutdown and placed in the cold condition when Wheeler Reservoir lake stage rises to a level such that water from the reservoir begins to run across the pumping station deck at elevation 565.Requirements for the instrumentation that monitors the reservoir level are given in Table 3.2.H.Descri tion of Re uirements: This Technical Specification has provisions for high reservoir water level instrumentation. A high reservoir water level indication is a preliminary indication of a flood.A flood is not a design basis accident or transient, thus reservoir water level is not credited in the safety analysis.Risk Justification: An analysis of the risk of external flooding was performed in BFN's Individual Plant Examination of External Events (IPEEE)for Severe Accident Vulnerabilities. Historical flood data was collected and analyzed to determine the frequency and magnitude of floods at BFN.All critical equipment essential to the safe shutdown of the plant are flood protected to an elevation well above that required in the current LCO.Given the plant design features and the conservative analysis of flooding in the IPEEE, the contribution of flooding to overall plant risk (probability of occurrence and radiological consequence) is considered negligible. Rl l b bill Rl l l ll l~kbk<<Low LowB-2 Enclosure V Volume 1S 3Pec 0'&on 3.'f, l NOV 18 1S88 4.6.D.R V v 588 3~+A~capon 4<<kz<gcs*< 8P<l ST 5 Z.g.2,~ggqg 3.The integrity of the relief valve bellovs shall be continuously monitored when valves incorporating the bcllovs design are installed. 3 6 E 2aMmua 4.ht least one relief valve shall be disassembled and inspected each operating cycle.E~Rum'.Whenever the reactor is in the SThRTUP or RUN modes, all)et yumya shall be operable.If it is determined that a)et pump is inoperable, or if tvo or more get pump flov instrument failures occur and cannot be corrected vithin 12 hours, an orderly shutdown shall be initiated and the reactor shall be placed in thc COLD'SHUTDOWN COHDITIOK vithin 24 hours.1.Whenever there is recirculation flov vith the reactor in tha STARTUP or RUE aodaa vith both recirculation pumps running, 5et yap operability ahall ba checked daily by verifying that the folloving conditions do not occur al simultaneously: ~ased log, VC'fig r SRZV.(,l~SR.,I 4~va racir at>~loo taaa-e-flo~~~~or<z.Ius aoea when'-i%~opera" b.Ink~I~l~The indicated value of core flov rata varies from the value derived froa looy flov measurements by sore than 10K.BFl Unit 1 3.6/4.6-11 c.The.diffuser to lover plan~di f ferantial pressure reading on an individual)ct yump varies from the mean of all]ct yump differentia.'ressures by aors than 10K.AMBINENT IjL Q~ 'i AUG 0 4 5$~<<5'uS44'r'Cab'O~ P~gQ~S 4 BFd I STS Z,q,~Pr~d Qp~6$R g...)4.6.Z.Whenever there is recirculation flov vith the reactor in the SThRHJP or RVH Mode and one recirculation pump i is operating, the diffuser to lover plenum differential pressure shall be checked daily and the differential prcssure of an individual)et pump.in a loop shall not vary from the mean of all get pump differential pressures in that looy by more than 10X.i&03igel+m r 1.The reactor shall not be operated vith one recirculation looy out~r;oeie of eerrfoe for sore theo~boors Vith thc reactor operating, if oac recirculation loop is out of service, the ylant shall bc placed in a HOT SHUTDOWNS CONDITION vithia 24 hours unless~the loop is sooner returned to.scrvicc.s, II 2 Recirculation pump s eeds shall be chewed e at least ce pcr ay.2~Fol oviag c y th dis ge va sp ed p may css t e spec p is css atsd s aed opera on, e of e lov t be o ened of th faster 50K of its 3~rgC7'>od D Mhaa the rca is n in thc CUR aode, CTOR POWER OPERATION vith both rec rcu at on pumps out-of-service for uy to 12 hours is permitted. Dur ag such interval estart of the recirculation yumps is permitted, yrovided thc loop discharge temperature is vithia 7$'F of the saturation 3.Bcf rc star ing cit r re rculat on ump d ing CT0 PO R IO, cck og th loo dis e teapc tur and do~at ati t cra ure.BEE Unit 1 sec 3MHg'~h'o 4 Qe~gds Pr ggh)(5T5 g tf 3'/4.6-12 AMENDMENT NO.2 g 7 f k.oo QvioN temperature of e reactor vessel vatcr aa determined by d pressure.The tota elapsed time natural circulation and one pump operation must be no greater than 24 hours.4.The reactor shall not be operated vith both recirculation pumps out-of-service vhile the reactor ia in the RUB mode.Polloving a trip of both recirculation pumps vhile in the RUN mode, imnediatcly inftiate a manual reactor scram.3.6.C 4+6.C The structural integrity of hSNE Code Class 1, 2, and 3 equivalent components shall be maintained in accordance vith Specification 4.6.G throughout the life of the plant.a.Vith the structural integrity of any ESNE Code Class 1 equivalent component, vhich fa part of the primary ayatca, not conforming to the above requirements, restore the structural integrity of the affected component to vithin ita limit or maintain the reactor coolant system in either a Cold Shutdovn conditfon or less than 50 F above the ainfiime temperature requfred by HDT considerations, until each indication of a defect haa been investigated and evaluated. 1.Inservice inspection of ESNE Code Class 1, Class 2, and Class 3 components ahall be.performed in accordance vith Section XI of the hSNE Boiler'and Pressure Vessel Code and applicable Addenda aa required by 10 CFR 50, Section 50.55a(g)j except vhsre specific vritten relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.hdditional inspections shall be performed on certain circumferential pipe vclda to provide additional protection against pipe vhip, vhich could damage auxiliary and control systems.BFS Unit 1 3.6/4.6-13 AMEKOMHP go p 06 0 S'F'.inca. 'on 3.M HAY 3 1594<CO Z.Q.)l.the reactor shall not be operated at a thermal pover and core flov inside of Regions I and II of Figure 3.5.N-l.2.If Region I of Figurc 3.5.N-1 is entered, faaedf ately initiate a aanual scram.SR Verify that the reactor is outside of Region I and II of Figure 3.5.5-1I a.Folloving any increase of aors than 5Z rated , theraal pover vhile initfal core flov fs less than 45K of rated, and 4An 8 If Region II of Figure 3.5.8-1 is entered: a.Iaecdfately initiate action and exit thc re ion vithin 2 u ert rol ods or by in reasi c c flo (sta ting eci culat on p t t ere fon an a 0 fa e ac 00)LR~b.While exiting the region, iaaediately initiate a manual acre%if thcraal-,hydraulf c instability is obserred, as cridcnced by o illa iona ch ed 1 percen pe to-p of r ted or LPRN acfll tions ch exes 30-pc ent peak-t peak seal.If peri ic LP ups e.o Qqps e ala oc ,'atel che the APRH's and indi idual RN's r erid ce of.the-hydraulic tab ity.b.Folloving any decrease of aors than 10K rated core flov vhile initial thermal pover is greater than 40K of rated.Pcopsr d Qc+'o n BlK Unit 1 3.5/4.5-21 NENbggPN.2p g

r~~~ill IIIIII I.~o UNIT 2 CURRENT TECHNICAL SP ECIF ICATION MARKUP 0 4 S S 0 5 cci figqgjo~g g/HOV 18 1988-'NITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS+L 4~sfili(qfio ~4r No~g~gg~Bylaw ISIS g yZ 3.g 3 3~4~The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. At least one relief valv shall be disassembled and inspected each operating cycle.3.6.E.~Jt Pupas E.~Jet um Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is determined that a jet pump is inoperable, or if two or more jet pump flow instrument failures occur and cannot be corrected within 12 hours, an orderly shutdown shall be initiated and the reactor shall be shutdown in the COLD SHUTDOWN COHDITIOH within 24 hours.~~5~ti 4/.Whenever there is recirculation flow with the reactor in the STARTUP or RUH modes with both recirculation pumps running, jet pump operability shall be checked daily by verifying that the following conditions p~~~g do not occur*-sa~.V.l.i simultaneousl ~v'cr.C'Lb a.The%vo ec rcu ation loopg%ave-a flow~:s~gt;>or ii ge kgb~.when.~opera~he-The dicated value of core flow rate varies from the value derived from loop flow measurements by more than 10X.h c,e J.4,!OOP c.The diffuser to lower plenun differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than lOX.BFH Unit 2 3 6/4.6-11 hMMmetn.%la

A'l AUG 04594 3854AL40~fol+a Q phJ Ql.l Pape@Nk~sR p.g.f.i 4'.E.2~Whenever there is recirculation flov vith the reactor in the SThRTtJP or RUR Mode and one recirculation pump ia operating, the diffuser to lover plenum differential pressure shall bc checked daily and the differential pressure of an individual)et pump in a loop shall not vary from the mean of all)et pump differential pressures in that loop by more lOX.QAI LCO 8.4)a4.44 Qo~~~:~~RZ.1.The reactor shal not be operated vith one recirculation loo out of service for more 24 hours.With the reactor operating, one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWR COHDITIOR vithin 24 hours unless the loop ia sooner returned to service.~0~9.V.//Recirculation pump ayceda M2.shall be checked A>at least once per A3 Al Pollov ng one pump operation, the d achargc va e of the v ape yump may t be open sa thc sp d of the f ter y ia less SO%of its eated speed.30 QVIOQ.D,,'Whan tha reactor ia not in the RJJK mode GTOR POWE HDtATIOR o recircu-on yea out~f-service for uy to 12 hours ia crmitted.au interval, restart o thc racirculation pumps ia permitted, provided the loop diachare temperature ia vithin'5 2'f the saturation temperature of the reactor 3~R2 Bcfor starting, either Foci ation dur REi POWER 0 OE, ck 1 thc lo y dia gc tcmyerat c and d saturation temperature ~BtÃUnit 2 c 5th W~d'IfiCagi~ ~~<C4~gg 4~BF<IsM z.9 9 3~6/4.6-12

Z eE<<$,.>.V.I NR i 8 1993 3~~gcgm8 h vessel water as determined by dome pressure The total e apsed time in natural circulation and one pump+I operation must be no greater than 24 hours.Secs.~);C;,.4;og. CE 4a<SftJ Lszg 4.pgsiod E The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.Following a trip of both recirculation pumps while in the RUN mode, ismediately initiate a manual reactor scram..G'.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to thc above requirements, restore the structural integrity of thc affcctcd component to within its limit or maintain thc reactor coolant system in either a COLD SHUTDOWN CONDITION or less than 50'F above thc minimum temperature required by NDT consider-ations, until each indication of a defect hae been inves<<tigatcd and evaluated. Inservice inspect,ion of ASIDE Code Class 1, Class 2, and Class 3 components shall be performed in accordance witt Section XI of the ASME Boil~and Pressure Veseecl Code anc applicable Addenda as requi: by 10 CFR 50, Section 50.55c except where specific writtc relief has been granted by 1 pursuant to 10 CFR.50, Sect: 50.55a(g)(6)(i). 2.Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.BFÃUnit 2 Se~~iS4eyio-4r C~~-gez*~cw5 Z.4.p//Q g i)~"'~Sec~ii<3 6/4~6-13 AMENDMENT lE 8 0 6 PAGE

Ai FEB 2 4 1995 L.t o L.l.Whenever the core thermal pover is g 25Z of rated, thc'ratio of FRP/CMFLPD shall bc Z 1.0, or thc APRM scram sctpoint equation listed in Section 2.1.k and the APRM rod block setpoint equation listed in the CORE OPEKLTIHG LIMITS REPORT shall be multiplied by FRP/CMFLPD. FRP/CMFLPD shall be dctermincd daily vhen the reactor is g 25Z of rated thermal pover.Qg J~g k Cite,ly~fi r CA'<7Q 4~8~N isis z.z.f 2.Shen it is determined that 3.5.L.1 is not being, mct, 6 hours is allovcd to correct the condition. 3.If 3.5.L.1 and 3.5.L.2 cannot bc mct, the reactor povcr shall be reduced to g 25X of rated thermal pover vithin 4 hours.Ai/CO 1.Thc reactor shall not bc operated at a thermal povcr and core flov inside of Regions I and II of Figure 3.5.N-1.2.If Region I of Figure 3.5.N-1 A is entcrcd, immediately initiate a manual scram.Sg'.0.I-2.1.Verify that the reactor is outside of Region I and II of Figure 3.5.N-1: a.Follovtng any increase of more than SX rated thermal povcr vhile'nitial core flow is less than 45X of rated, and 3.If Ze~g II of.Figurc 3.5.N-1 is'entered: b.Follovtng any decrease of more than lOX rated core flow vhilc initial thermal povcr is greater than 40X of rated.BFH Unit 2 3'/4.5-20 aemoMapgo. 2S2 5',- N OO 3~~~a.yq IG~8 Immediately initiate action and exit the region vithin 2 hours nsert ng con ro o s or b increas ng cor f v.('tar g a re rcu-1st~n pump exit t region s~ot an appropriate action)and b.While exiting the region, immediately initiate a manual scram if thermal-hydraulic instabilit is obse as evidenc d by AP oscil a-ti ns vh ch excee 10 p cent pea-to-p ak of r ed or PRN osci atio vhich exceed 30 pe ent eak-to-eak of cale.'Cf p riodic LP3X scale r d vnscale alarms oc r, edi ely ch ck the APRM and ndi idual ARM's for e denc of thermal-hydraulic ins ability.BFN Unit 2 3;5/4.5-20a

100.3.V.(-(Fi ure BEN Power I=low Stabilii:y Regions Q~00000 9~~0 000~0~0~0~~0~0~~100%Rod Line.R 0 CJ O I 0 C 6)O 0)CL ID 0 CL 0)0 (3 80-70-~~0~~0~~0~000 f 000~000~g~0~0~0~0~~\~0~~0~~0~50.~~~~40-.30--.Noturol (:irculotion, Line Q 00 0~~~00~~~~2 10-~~~~oo~0~~~~~~~~00~0~~0 0 PP~I 5 10 01~tW 15 20 25 30 Note: OperoVion Not 60;----Permitled in TI>is Region 807.Rod Line~~~eend~~~~~~\~~~~~~~~~~~~0~0~~~~~~0 l[jggm Raijion Rnoion Io~~~~~~~~~~~~~~r~~rm r~<~~35 40 45 50 55 60 65/0/5 80 85 90 95 100 10 Core Flow (perrcnt of rnted)CD C3 CTl 5 0' UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP NG CO TONS FO 0 ON SPeC Plea.OY 18 tS88 SUR C See 3WS+06CC fjOn 4e~eS+r tlPA 1sT5 s.e.x~30 9e 3.The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. 4.At least one relief valve shall be disassembled and inspected each operating cycle.3.6.E.J~ee Pum E.J~e~im~1.Whenever the reactor is in the STARTUP or RUN modes, all jet pumps shall be OPERABLE.If it is determined that a jet pum'p is INOPERABLE, or if two or more jet pump flow instrument failures occur and cannot be corrected within 12 hours, an orderly shutdown shall be initiated and the reactor shall!be placed in the COLD SHUTDOWN CONDITION within 24 hours.583.'].I-I Whenever there is recirculation flow with the reactor in the STARTUP or RUN modes with both recirculation pumps running, jet pump operability shall be checked daily by verifying that the following conditions do not occur simultaneously: +~'y a X~vo ecirculati oops hes~a flow~(s]of or 4504K.when l]4', 1 6C~opera tC'E.E nu+lc El b.The indicated value of core flow rate varies from the value derived from loop flow measure-ments by more than lOX.0 BFN Unit 3 3.6/4.6-11 c.The diffuser to lower plenum differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10K.AIAENDINBII'5. 1 9 9~~a=..& 0 h BO ARY Ai QPgc'gjcctM'1 3b/~(AtJS 04594 5Ll564i4AW oA 6<Cl~es e Bknl t S TS Z.9-'2-L f TO)05ed@ok 6 5R 3A.I.I 4 6 E.~Jam I 2.Whene ver there is recirculation flow with the reactor in the STARTUP or RUH Mode and one recirculation pump is operating, the diffuser to lower plen~differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all get pump differential pressures in that loop by more than 10K.i co z.e.(R4c eeerrrrm rrrr~lrr2 1.The reactor shall not be operated with one recirculation loop out of service for more than 4 hours.With the reactor operating, if C+g one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH within 24 hours unless the loop is sooner returned to service.SR3.~I l 1.Recirculation pump speeds shall be checked'm 1 at east nce er a Laz.2.~LA I F liow ng e-p p o era ion, e sch ge alve of e 1 pe p m no be pene un ss e eed f e f te p p i le s t 5 of ts ate spe d.3.ACTIN al D When th eactor is not-'in th RUH mode, REACTOR POWER OPERATIO with both recirculation pumps out-of-service for u to 12 hours is permitted. uring such interval estart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature 3.Bef e st rti eit er re rcu tion pum d ing CT P R ERA OH, e and og e lo p d cha ge tern ratu e do e saturation tempe ture.BFN Unit 3 See yooiiCrcrHon for<largP 6" BFN~SYS Z.H.9 0 Agio g D of the reactor vessel water determined b dome ress The tota e apsed time fn natural circulation and one pump operation must be no greater than 24 hours.See 5~sWWog fc~~~<krBPN igTsp q.q, 4~AC4'oui, E The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G S ctu 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1"equivalent component, which is part of the primary system, not conforming to the above r'equirements, restore the structural fntegrity of the affected component to within fts limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F above the miniinm temperature required by HDT consfder-ations, until each indication of a defect has been investigated and evaluated. l.Inservice inspection of ASME Code Class 1, Class 2, and Class 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, Section 50.55a(g except where specific written relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additional inspections shall b performed on certain circumferential pipe welds to provide additional protection against pipe whip which could damage auxiliary and control systems.BFH Unit 3 See S~W'cOHo+6~ch~ys 6 cps s.<%6.r..G in yh;S geC4og 3.6/4.6-13 ANBMgr NO.y 79 paG 0 ~~~~;~:w,~ats&e1iJBYilt(hsrtit&eloeily7%(lMJt Ram~qn w;I~P$:~.'i+igl J g4'A>4f yy~~XI)'EMIIHIIT') CtIi1'J Sa'e->~~~~'~~~~~~~~~~~~~~~~~~~~~'Q'~~~~~~I~I~~~~I~~~~~~~o~.~g~~,~I~I~I~~~~~~~~I~~~'I~~~'~~~I~~4t~~.I~~~~~~~~~~~I~~~~~~~~I~~~~~I a~~I~~~>~~I~~~1~~~~~~~~~~~'I>~~~~i~~

~~~I~~~~~ E JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in'a technical change.CTS requires the plant to be placed in the HOT SHUTDOWN CONDITION in 24 hours with one recirculation loop out of service.Proposed ACTION C requires the loop be returned to service in 12 hours or ACTION D requires the plant to be in MODE 3 (Hot Shutdown)in 12 hours.The CTS and the proposed ISTS Completion Times are essentially equivalent since both require the plant to be in MODE 3 in 24 hours.A3 The frequency for this Surveillance has been changed from once per day to once per 24 hours.This is a terminology change and is therefore administrative. TECHNICAL CHANGE-MORE RESTRICTIVE CTS allows up to 24 hours operation with the reactor power<1%with no recirculation loops operating (the total elapsed time in natural circulation and one pump operation must be no greater than 24 hours).Proposed ACTION D is more restrictive since the time limit of 12 hours applies to<1%while in MODE 2 also.0 BFN-UNITS 1, 2, 5.3 Revision 0 Pi,GE Vt 3 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING H2 The flow imbalance limit is being reduced to 10%of rated core flow when operating at<70%of rated core flow, and to 5%of rated core flow when operating at a 70%of rated core flow.The current requirement is 15%mismatch of flow at the given flow conditions. While the limit appears to be less restrictive if core flow is x 66%of rated core flow, it is more restrictive when'>66%of rated core flow (i.e., 15%x 66%or less is x 10%of rated core flow), where the unit normally operates.In addition, currently, this is only a problem if there is an imbalance in combination with two other conditions (CTS 4.6.8.l.b and c).The new requirement is separate from the other two, thus, actions will now be required if there is an imbalance by itself.Therefore, this change is considered more restrictive on plant operations. TECHNICAL CHANGE-LESS RESTRICTIVE"Generic"~LA1 This requirement is being relocated to plant specific procedures. The purpose of this limitation is" to provide assurance that when shifting from one to two loop operations, excessive vibration of the jet pump risers will not occur.Short term excessive vibration should not result in immediate inoperability of a jet pump, but could reduce the lifetime of the jet pump.This type of requirement is generally found in plant operating procedures, similar to other operating requirements necessary to minimize the potential of damage to components. Changes to the procedures will be controlled by the licensee controlled programs.LA2 This requirement is being relocated to plant specific procedures. Details of the methods for performing this Surveillance, and any requirement to record data, has been relocated to plant procedures. Any changes to the procedures will be controlled by the licensee controlled programs.LA3 These requirements are being relocated to plant specific procedures. The details of the acceptable method for meeting an action requirement and what constitutes evidence of thermal hydraulic instability and the need to check for it have been relocated to plant procedures. Any changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 5 3 PAGE~OP Revision 0 Cl JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING"Specific" Ll This change adds a note which states the Surveillance is not required to be performed until 24 hours after both recirculation loops are in operation. The Surveillance is not required to be performed until both loops are in operation since the mismatch limits are meaningless during single loop'or natural circulation operation. Also, the Surveillance is allowed to be delayed 24 hours after both recirculation loops are in operation. This allows time to establish appropriate conditions for the test to be performed. L2 Per CTS 3.5.M.3.a, if Region II of Figure 3.5.M-1 is not exited within 2 hours, the Specification is violated and CTS 1.O.C.1 applies requiring the plant be placed in Hot Standby within 6 hours and in Cold Shutdown within the following 30 hours.This provides actions for circumstances not directly provided for in the specifications and where occurrence would violate the intent of the specification. The BFN ISTS provides Action within the Specification which could be considered less restrictive than CTS.Action 0 allows 12 hours to be in MODE 3 (Hot Shutdown)and 36 hours to be in MODE 4 (Cold Shutdown). The proposed Action is considered less restrictive since 12 hours is allowed to place the unit in Hot Shutdown versus the 6 hours allowed to place the unit in Hot Standby per CTS.BFN-UNITS 1, 2,&3 Revision 0 UNIT 1 CURRENT TECHNICAL SP ECIF ICATION MARKUP NOV Z8 1988 4.6.D.R e Va ve 5'C'('~c+',4;(c,h~ p+~$<o+<BP+ILATS 3,q, 3~The integrity of the relief valve bcllovs shall be continuously monitored vhen valves incorporating the belloys design arc installed. LCo 3.9.2 Whenever thc reactor is in thc Rgplicab'1'kj STARTUP or RUH modes, all,)et umps shall be operable.If it is determined that a jet pump is inoperablc, r tv or re t pump ov t cnt fa urea occur d c ot c rec d 12 ours an order y shutdovn shall bc initiated and the reactor shall'be placed in the~HUTDOWH COHDITIOH vithin ours.4~At least onc relief valve shall be disassembled and inspected each o erati c cle.Vcn c&oLht.o~(. of-lt~~i~P C~ehn~Mti Fi'CA (eLcg o~ee 0 A iowan enevcr there is ecircu ation flov vith he re ctor th TAR or cs ith 0th ecir Lt um s ingg ct op rabil s 11 b checked aily vcrifyi t th follov ng c diti do not occ simultaneously: ~e Y~ShA(cfhon 4,f-~~>gci gPN LS75 p,q.~frogoCd<R>'t>I IJokS g2 ftopsH SP Z.9.3.I a.The tvo recirculation loops have a flov imbalance of 15X or morc vhen the pumps arc op~rated at th s ccd b.The indic ted value of orc ov ate v ies rom e luc eriv f oop ov mess em ts more han 10K eke cc~]isa&%Pe v l.$.The dfffeeer eo lower plenum dif ferential pressure reading on an individual)et p varies from BFH Unit 1 3.6/4.6-11 C g less g2.NENOMENT NL I 5 8

Se>>iP;on 5.,z AUB 04 1994 3.6.F~>m Z~Von 4-C/g~~k~8pnl Isis z.g,~~se~lICab;fig 4.6.F ev r t ereyis cu ti A,o vith thc reactor in the~STARTUP or RUR Mode and n rec culat n um 8 diffuser to lover plenum differential pressure shall be checked a and thc differential pressure of an individual get pump.in a loop shall vary from m of all get p dif er tial (mesa rea t t o by eave~z~ass t}x 1.The reactor shall not bc operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if onc recirculation loop is out of service, thc plant shall bc placed in a HOT SHUTDOWN CORDITIOR vithin 24 hours unless the loop is sooner returned to service.1.Recirculation pump speeds shall be checked a'nd logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed pump may not bc opened unlesa the speed of thc faster pump is less than 50Z of its.rated speed.3.When the reactor%a not in thc RUE mode, REACTOR POWER OPERATIOR vith both recirculation pumps out-of-service for up to 12 hours is permitted. During such interval restart of the recirculation pumps ia permitted, provided thc loop discharge temperature is vithin 75'P of the satu 2.Ro additional surveillance required.3.Before starting either recirculation pump during REACTOR POWER OPERATIOK, check and log thc loop discharge temperature and dome saturation temperature. BFK Unit 1 3.6/4.6-12 AMENDMENT No.2 g y-:.3-

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~GP~ 0 ION s NOV 18 1S88 4.6.D.e Va ves 5'VC7 lgjCA7 IohJ Fog~ho/~/or B j=w 3~The integrity of the relief valve bellovs shall be continuously monitored vhen valves incorporating the bellows design are installed. Al~~~Lu)3.VZ I-~Apl'c4'.j,g Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is determined that a Jet pump is inoperable or i~two or ore get pump flov instibment failu s occur~cannot orrecte vithin 12 hours an orderly shutdown shall be initiated and the reactor shall be shutdown in the 66BB-HUTDOWH COHDITIOH vithi hou/2.ad QuST(@CA Tkr&FbA C4<u~84, BFH'ggg g g.]4~At least one relief valve shall be disassembled and inspected each'erati c cle.c Q ak o<c.C1.4'ra lS Sa'l<gg~CI h'4 i~g/enever t vk~/os recircul ion flov vith the re tor in the TAR or RUH modes it both rec culation p ps runni, get pump operabili shall be checked aily b veriiyi tha the follovfng c ditions do not occur a.The tvo recirculation loops have a flov imbalance of 1SX or more vhen the pumps are operated at the same s eed.4~~~SW Sa S.MZ.I No4S~b.The i icated value of c re flov ate var es from e.v ue deri ed ir m oop flo measurements by more BFH Unit 2 Prolog~Sg.3.Q 2./~dc SR 3.42.y$~~as'I~4/p ffcr~3.6/4.6-11 The diffuser to lover plena differential pressure reading on an individual et pump varie from by p./'MENDMEÃf IL I 54 Al 3.6.F SeeadÃc4m 4~a fc~BFN Is~~'I.J t 0 t D~sg 3.Cz,l Wh never there is reci c atioWflo vith]the reactor in the"'"'~M (STARXUP or RUH mode~one~ec~u at o~uidp o erati , t c diffuser to lover plenum differential pressure shall be checked pS@~ail and the differential pressure of an individual jet pump in a loop shall vary from e me o all)et ump-di ential essurc loop y than~C5 2d L3 4.6.F.1.The reactor shall not be operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if one recirculation loop is out of service, thc plant shall be placed in a HOT SHUTDOWH COHDITIOH vithin 24 hours unless the loop is sooner returned to service+1.Recirculation pump speeds shall be checked and logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed pump may not bc opened unless thc spccd of the faster pump is less than 50K of its rated speed.2.Ho additional surveillance required.3.When thc reacHmis not-in the RUE mode, REACTOR POWER OPERATIOH vith both recircu-lation pumps out-of-service for up to 12 hours is permitted. During such interval, restart of the recirculation pumps is permitted, provided the loop discharge tempcraturc.is vithin 75 F of thc saturation temperature of the reactor BFH Unit 2 3.6/4.6-12 3.Before starting cithcr recirculation pump during REACTOR POWER OPERATIOH, check and log the loop discharge temperature and dome saturation tcmpcraturc. AMENOMENr N.2 2 g.3..-3-- 0, UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 0! 5 e'A'cn',g.g HOV18 1S88 4.6.D.Re e Va ves See~~ggg~+~Shd IS TS yq.g 3~The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. Ecole,g 2.enever the reactor is in the TARTUP or RUH modes, all jet~pumps shall be OPERABLE.If t is determined that a jet pump is INOPERABLE or or et pf wist nt fai ure occu and anno be orr cte with s an order y s utdown shall be initiated and the reactor shall be placed in the HUTDOWH CONDITION within ou Iz Ito7.~de+reeH~ifrrfroe gr Cluepr+~8PN.Is<5 g,q.i QA>4.At least one relief valve shall be disassembled and inspected each operating cycle.gtt 4 lcf36+o~6))auo'f~enev r recir ulati n fl w t the eact in he ST TUP R mo es w th bo re rc at n umps i , et p oper ilit s 11 che ed ily by veifyi t t t f liow ndi o o no oc r simultaneousl '.The two recirculatio loops have a flow imbalance of 15K or more when the pumps are operated at the same speed.J-Z Ifogsbcd 5R 3.9.g,(Plpkg b.e n at d lu of or fl w at v ie fr m he lu d iv d f om oo f w eas re-m ts y or than 1 X.sa e.~,>>The diffeeer re lever plenum differential pressure reading on an individual jet pump varies fire++ho flC csW0Q4e fh Hrea BFH Unit 3 3.6/4.6-11 than~40'/o la>>AMENOMHfT Ni)f 29,;P/s+t-~ ~~Ccs c~~AUS 0 4 594 5<+YuSkjkicaHon h r t:kpeg+~8~l SVS 8 N.1~~~SR cn er her is rc at on ov vith Ay~;cog],'Q~) the reactor in he C STARTUP or RUH Mo and e c ula i o c ati the di fuser to lover plenty differential pressure shall be che ail and the differential prcssure of an individual jet pump in a loop shall/}C va fro t m of a et ump h3 if cr tial s e t t o by move than 3.6.F e c at 0 a o 4.6.F t o 1.Thc reactor shall not be operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH within 24 hours unless thc loop is sooner returned to service.1.Recirculation pump speeds shall be checked and logged at least once per day.2.Folloving one-pump operation, the discharge valve of thc lov speed pump may not be opened unless thc speed of the faster pump is less than 50X of its rated speed.2.Ho additional surveillance required.3.When the reactor is not"in the RUH mode, REACTOR POWER OPERATIOH vith both recirculation pumps out-of-service for up to 12 hours is permitted. During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is vithin 75'F of the saturation temperature 3.Before starting either recirculation pump during REACTOR POWER OPERATIOH, check and log thc loop discharge temperature and dome saturation temperature. BFH Unit 3 3.6/4.6-12 a~arn~mt vo.i 8C JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUNPS ADN I NI STRATI VE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readil'y readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change..The wording of the surveillance was changed to require verification that one of the following criteria are met rather than verifying that none of the conditions exist simultaneously. This is consistent with NUREG-1433 which attempts to phrase everything in a positive manner.Due to the change in phrasing of the Surveillance,"more than" was changed to"less than or equal to" in criteria b and c.A3 The variance of the diffuser-to-lower plenum differential pressure reading on an individual jet pump will now be taken from the established pattern rather than from the mean of all jet pump differential pressures. This change is in accordance with the recommendations of SIL-330 and NUREG/CR-3052 and is consistent with NUREG-1433. A4 The conditions of the Surveillance Requirement are assured by LCO 3.4.1.Therefore, there is no need to restate the conditions for jet pump operability. A5 The frequency for this Surveillance has been changed from daily to once per 24 hours.This is a terminology change and is therefore administrative'. BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUMPS TECHNICAL CHANGE-MORE RESTRICTIVE Ml The requirement to place the plant in a Cold Shutdown condition within 24 hours when a jet pump is inoperable has been revised to reflect placing the plant in a non-applicable condition. Current Specification 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in Cold Shutdown is not applicable after Mode 3 is reached.The'revised action requires plant power to be brought to Mode 3 (outside the applicable condition) within 12 hours.The current action allows 24 hours to place'he plant in a non-applicable condition. As such, this is an additional restriction on plant operation which constitutes a more restrictive change.This change adds two requirements to the Surveillance to detect significant degradation in jet pump performance that precedes jet pump failure.The first requirement added would detect a change in the relationship between pump speed, and pump flow and loop flow (difference >5%).A change in the relationship indicates a plug flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle.The second requirement added monitors the jet pump flow versus established patterns.Any deviations >10%from normal are considered indicative of potential problem in the recirculation drive flow or jet pump system.These two added requirements to the Surveillance help to detect significant degradation in jet pump performance that precedes jet pump failure.Requirements added to Surveillance Requirements constitute a more restrictive change.In addition, CTS 4.6.E.1 allows jet pump operability to be verified by demonstrating that the two recirculation loops.have a flow imbalance of s 15%when the pumps are operated at the same speed.This is now a separate requirement (Proposed SR 3.4.1.1 See M2 of the Justification for Changes for Specification 3.4.1)and can no longer be used by itself to demonstrate jet pump operability. This change is consistent with NUREG-1433. 0 SIL-330 provides two alternate testing criteria (thus the deletion of current Surveillance 4.6.E.l.b).One method uses easy to perform surveillances with strict limits to initially screen jet pump operability (the proposed changes above).If these limits are not met, another set of Surveillances exist (current Technical Specifications). Revising the Surveillances to separate the flow imbalance test requirement and to include the stricter limits reflects a more restrictive change.BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUMPS TECHNICAL CHANGE-LESS RESTRICTIVE"Specific" Ll This change deletes the current shutdown requirement associated with jet pump flow indication. Currently, when required jet pump flow indication is lost, an orderly shutdown must be initiated in 12 hours and the reactor is required to be in Cold Shutdown within the following 24 hours (since Mode 3 is the non-applicable mode, then 24 hours is allowed to reach Mode 3;see discussion of change Hl for ITS 3.4.2).The proposed Specification implicitly requires the jet pump flow indication to be operable only for the performance of the Surveillance Requirement. If the flow indication is inoperable when the surveillance is required to be performed and jet pump flow can not be determined by other means, the jet pump would be decl'ared inoperable and the appropriate actions would be followed.Since the proposed jet pump surveillance requirement is required to be performed every 24 hours (the 25%extension per SR 3.0.2 can be applied)and the Required Actions require the reactor to be in Mode 3 within 12 hours, the maximum difference in the current Specification and the proposed specification is 6 hours.As a result, the proposed specification effectively allows a maximum of an additional 6 hours (which is the 25%extension) to reach a non-applicable Mode if a required core flow indicator is inoperable and jet pump flow can not be determined. Depending on when the failure occurs, 6 hours is the maximum increase over the current Specifications (failure occurring immediately after the surveillance is performed). The following table provides the details of the calculation of the 6 hour period: Current Tech Specs Time 0 hours-Jet Pump.Indication Fails-12 hr AOT Begins Time 12 hours-12 hr AOT Expires-24 hr AOT Begins to MODE 3 (per 3.0.A;see Ml)Time 36 hours-24 hr AOT Expires Plant in MODE 3 Proposed Tech Specs Time 0 hours-Jet Pump Indication Fails (Immedi ately After SR Time 30 hours-SR due;Flow (24 hrs x Indication Inop 1.25)-12 hr AOT to MODE 3 Begins Time 42 hours-12 hour AOT Expires Plant in MODE 3 BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR.CHANGES BFN ISTS 3.4.2-JET PUMPS L2 As depicted above, 42 hours is the maximum time that would be allowed if a required jet pump flow indicator is inoperable and jet pump flow can not be determined. Currently a maximum of 36 hours is allowed if more than one jet pump flow indicator is inoperable. Jet pump flow indication operability does not directly impact jet pump operability. Jet pump flow indication is only required to perform the jet pump Surveillance (SR 3..4.2.1).SR 3.4.2.1 verifies jet pump operability and has a frequency of every 24 hours.The 24 hours frequency plus the 25%extension has been shown by operating experience to.be timely for detecting jet pump degradation and is consistent with the surveillance frequency for recirculation loop operability verification. The most common outcome.of the performance of a surveillance is the successful demonstration that the acceptance criteria are satisfied. This change is consistent with NUREG-1433. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation. Note 2 to proposed SR 3.4.2.1 provides time to perform the required.Surveillance when the reactor exceeds 25%RTP.Below 25%RTP, low jet pump flow results in indication which precludes the collection of repeatable and meaningful data.The flexibility to proceed to a 25%RTP and then commence the SR every 24 hours is consistent with approved Technical Specifications for both Perry Nuclear Power Plant and River Bend Station.L3 The allowed difference between each jet pump diffuser-to-lower plenum differential pressure to the loop average has been increased to 20%.This change is consistent with the recommendations of SIL'-330 and NUREG/CR-3052 (Closeout of IE Bulletin 80-07: BWR Jet Pump Assembly Failure).SIL-330 specifies a 10/criteria for individual jet pump flow distribution. When measured by jet pump diffuser-to-lower plenum differential pressure, the equivalent limit is 20%because of the relationship between flow and delta-P.Since BFN uses the diffuser-to-lower plenum differential pressure measurement, the variance allowed should be 20%as recommended by SIL-330 and NUREG/CR-3052. This is a relaxation from existing requirements, therefore, it constitutes a less restrictive change.This increase in allowed difference is considered an acceptable criterion for verifying jet pump operability and is consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 PAGE

UNIT 1 CURRENT TECHNICAL SPECIFICATION .MARKUP

OEC 0 7 1SS4 F 6.C 4'.C 2~Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours I for the air sampling system.,!2.With the air sampling sys tern inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.Ac 5'usaf'i~tjon gag Qqg HPFA)ST>pgq y~~~The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.3~Qai 1.A<Ho 8 R L2 If the condition in 1 or 2 above cannot be met, an orderly shutdoml shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.I 4Me3 in AH I>br s When o relief alve is kno~to be failed, an orderly shutdo~shall be i iti and the reactor epressurize o less than 105 w th hour e s ae ot eq red to b OP LE in e C IT N.SR 3.ge3ol A proxy e y one-half of all re ief va es sha 1 be b ch-ch ked or r laced with a eck ve A 2 each operatin c cle.All val es 11 have en heck or r lac d up the co let on of ever ecnd c es l,243 tfopScl SR s.q',g,~SR7'l 3~2 In accordance vith Specification MM each relief valve shall be manually opened t'1 t e oco ples nd ac stic moni rs d str am of the alve i dica e ste is lookin f the a lve.BFN Unit 1 3.6/4.6-10 AMENDMENT NL 2 Z 3 0 gkl NOY 18 1988 3.e i tegr y of the elie val e bel ovs shal be ontin ousl mo tore vhen alv in orpo ating he b 1 ws desi r tall d.At leas o rel ef al e s all e sass bl d d i p ted ach oper ti c e.3 6 E Jm~muu E.~Jet 1.Whenever the reactor is in the STARTUP or RUH modes, all get pumps shall be operable.If it is determined that a)et pump is inoperable, or if tvo or more Jet pump flow instrument failures occur and cannot be corrected vithin 12 hours, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.1.Whenever there is recirculation flov vith the reactor in the STARTUP or RUH modes vith both recirculation pumps running, get pump operability'hall be checked daily by verifying that the folloving conditions do not occur simultaneously: gee rw+,'p;(qp<g~<+es tr>8cnr lSTS~s.z.r+P~ps a.The tvo recirculation loops have a flov imbalance of 15Z or more vhen the pumps are operated at the same speed.b.The indicated value of core flov rate varies from the value derived from loop flov measurements by more than 1OZ.BFH Unit 1 3.6/4.6-11 c.The diffuser to lover plenum differential pressure reading on an individual)et pump varies from the mean of all)et pump differenti pressures by more than 10Z.AMENOMENr N.Z g 8 p~GC a.' 0 SAPBTY I INIT 1.2 Reactor Coolant S stea Inte rlt f INITIN3 SAPETY SYSTEN SBTTIHQ 2.2 Reactor Coolant S tea Inte rlt Applies to llalts on reactor coolant systea pressure.Applies to trip settings of thc instruecnts and devices which are provided to prevent the reactor systea safety lialts fraa being exceeded.O~b8cl 1v8 To establish a ligilt below which the integrity of the reactor coolant systea is not threatened due to an overpressure condition. o~h ective To define the level of the process variables at which autccaatic protective action is initiated to prevent the pressure safety limit free" being exceeded.S ecificatlons A.The pressure at the lowest point of the reactor vessel shall not exceed 1,375 psig whenever irradiated fuel is in the reactor vessel.Cgz~gc 4 gC 4 I S TS, 3.o The limiting safety systea settings shall be as specified below: Liwiting Safety rotcctlve Action S tea Settin SR 2'f.3l h.Nuclear systea 1.105 psig+relief valves sg, sl open-nuclear (4 valves)systca pressure 1,115 psig+33.5 osl (4 valves)~1.125 pslg+83.8-k+psl (5 valves)B.Scraa--nuclear <1,055 psig systen high pressure BFN unit l 1.2/2.2-1 S UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 0 OEt;0 V 1SS4 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout providing a temporary monitor.2.Pith the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.5+8~ST(FICA7/QPJ~Q CPA+G:E-svo BFN)'sl-s gc/g+3.0.5 p<glot4 p L,h goo&5 lg2Q)),ca),;l Q When ore than one relief valve is kaovn to be failed an orderly shutdovn shall b initiate e react r depress o lyas 10~ig vithin ours.e c ar ot req ed to bc in COLD SHUTDOWN 7rapos~Qp3 hble.ha sas4~~3.If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall be initiated and the reactor shall be placed in the COLD SHUTDOWN COlQITIOK vithin 24 hours.sp s..~l tely one-half o~l?,he al relief val s sh 11 be.ch-cheched r repla d vith a p,Z bench-chewed valv each operating cyc JQ.1 13 valves vill have h!Ien chere~or replace~on the compMtion of every 2.In accordance vith/,Z.5g.4'f.3.> Specification.O.tR each relief valve shall be aanually opened unti c c co tic m tora o trc of the alv dicate team i flo ng om thc valve.BFR 3 6/4.6-10 AMENDMENT 10.2 29 Fia=-~OF~ I HOY 18 1S88~r~~4 3 Tho r te'ty of che eh on be ont o ly to v es rpo atQxg the bellove design are install 4, leaa one r ief alv e di emb d azicl act ea the e.3.6A.&~max 1.Whenever the reactor is in the STOUP oz EHf modes, all Jet pampa ahall be OPELQKZ.it ia determined that a get pap is inoperable~ or if tvo or more)et pump flov instrument failures occur end cannot be corrected vithin 12 hours, sn orderly shntdovn shall be initiated and the reactor ahall be ekan~rn in the COLD SKFRtNN[-CanamOI vithin Zi~.QQ Qug4s ls qgL>0M C4c.~q*Q,Phd<gP~2 4 2, a e,k P-1 ps.1.Rxenever there is recirculation flov vith the reactor in the STQCUP or RON modes vith both recirculation poaye ramduS, get pmnp operability shall be checJcack daily by veri~that the f ollcndxw conditions do not occur simnltaneoasly1 a.The tvo recirculation loops have a flov imbalance of LSZ or more vhen the pampa are operated at the same speed+b.~indicated value of core flov rate varies from the value derived from loop flov measurements by more than 10K.c~The diffnaer to lover plenum differential pressure readfilg on an individuaJ, p1mep varies from the mean of all)et pump differential pressures by more than 10".Vn't" 3.h/4.6-11 AMENDMENT NO 15 4 SAFETY LIMIT l.2 Reactor Coolant S stem Inte rit LINITING SAFETY SYSTEN SETTING 2.2 Reactor Coolant S tern Inte rit Applies to limits on reactor coolant system prcssure'~e ective To establish a limit below which the integrity of the reactor coolant system is not threatened due to an overpressure condition. Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits frcci being exceeded.O~e votive To define the level of the process variables at which automatic protective action is initiated to prevent the pressure safety limit from being exceeded.. S ecifications S cifications A.The pressure at the lowest point of the reactor vessel shall not exceed 1,375 psig whenever irradiated fuel is in the reactor vessel.The limiting safety system settings shall be as specified below'ijiiting Safety protective Action S stem Settin 5R 9.0,9./h.Nuclear system 1.105 psig+relief valves 93 psi open-nuclear (1 valves)system pressure SEE'Q5TIF'Ic/TIoAJ F'ag CPANg gee~O gyral i~g Z~.115 psig+-3i.S (i valves)1,125 psig+si (5 valves)B.Scram-nuclear<1,055 psig system high pressure BPN Unit 2 1.2/2.2-1 PAGE UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP PAGE~OF~ 0 5 DEt'7 1994~6.C 4.6.C 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling'systems shall be OPERABLE'rom and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.2.With the air sampling system inoperable, grab samples shall be obtained and analyted at least once every 24 hours.~~X~564ie4on 0~chases+BIN jsTs 3'.q.y p3.V.s'he air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.od<$lp X43+l jCR bIll ProPOSC Jj HO<~SP.3.la.>3.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.'in M>jn jIhfS 1.When than on relief valve is known to be failed, an orderly shutdam shall b initiate d the reactor depressurite o lcaa 0 sx v thin~e LR e c s a n re i ed t e 0 ERAB i th CO CON TIO sg P.f>.sR z.s.s.c 2 proximately onc-half o 11 relief val s shal e bench-chec d or repl d vt,th a bench-ch eked vc, ach o r n c cle hll 13~alves vill have be checked r repla d upon t complet of eve second cycle.If'n accordance with Specification each relief valve shall be manual o ed t thermoco es and aco tic monito downs earn of the alve indicat steam is the valve BFH Unit 3 3.6/4.6-10 AMENVHrr NIL X 86 Po.GE~O'~ 0 SPecif)cafjnr) 3.Q.3'OV 18 t988 Q5 3.Th in egrity of the r ie val e b 11 vs~al bc ont nu usl on ore vh alv s inc rpo ati e ellovs de ign rc ns al d.4.At eas one rel ef v lve s ll di ass ble d i pcc ed ach 0 cra ing cyc~3.6.E.Jet~ups 1.Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is dctermincd that a jet pump is IHOPERABLE, or if tvo or more jet pump flov instrument failures occur and cannot be corrected vithin 12 hours, an.orderly shutdown shall be.initiated and thc reactor shall be placed in the COLD SHUTDOWN COHDITIOH vithin 24 hours.E.J~e 1.Whenever there is recirculation flov vith the reactor in the STARTUP or RUH modes vith both recirculation pumps running, jet pump operability shall be checked daily by verifying that the folloving conditions do not occur simultaneously: 5'ee Tus&ceh>>n J>>~c4ngc's 4 BcN 7srs g.q.2 5<fpuw p g a.The tvo rccirculatio loops have a flov imbalance of 15X or more vhen the pumps arc operated at the same speed>>b.The indicated value of core flov rate varies'from the value derived from loop flov measurc-mcnts by more than 1OX.BFH Unit 3 3.6/4.6-11 c.The diffuser to lover plenum differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10X.AMENOMBfr N.I 29 5fec)@ca an 3'.<3 1.2 2.2 hpplles to liaits on reactor coolant systea prcssure.Applies to trip settings of thc instruments and devices vhich are provided to prevent thc reactor systca safety liaits froa being excccded.To establish a liait belov vhich thc integrity of thc reactor coolant systea ls not thrcatencd due to an ovcrpressure condition. To define thc level of thc process variables at vhich automatic protective action is initiated to prcvcnt thc pressure safety lialt froa being exceeded.h.The prcssure at thc lovcst point of the reactor vessel shall not exceed 1,375 psig vhenevcr irradiated fuel is ln thc reactor vcsselo The liaiting safety systea~ettings ahall bc as specified bclov!5R S4,.I J.Nuclear systea 1 105 pslg g relief valves 33,xM psi open-nuclear" (4 valves)systea pressure 115 craig g'f m S psl 4 valves)Scc SLL5+s gasw 4s f5'2eg 1 125 psig g.g~psi (5 valves)B.Scraa-nuclear g1,055 psig systea high prcssure BPK 1.2/2.2-1 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF VALVES ADMINISTRATIVE A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The Frequency for proposed SR 3.4.3.1 (CTS 4.6.0.1)has been changed from"each operating cycle" to"18 months." Since an operating cycle is 18 months these are equivalent. The Frequency for proposed SR 3.4.3.2 (CTS 4.6.0.2)has been changed from"In accordance with Specification 1.0 HH" to"18 months." Since the Inservice Testing Program (1.0.HH)frequency is 18 months these are equivalent. As such, these changes are considered administrative. A3 The proposed change adds a note that states that the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASHE code requirements, prior to valve installation. As such, the addition of the note is considered administrative. A4 CTS 3.6.D.1 requires an orderly shutdown when more than one relief valve is known to have failed.Therefore, the CTS allows unlimited operation with one S/RV inoperable. BFN has 13'/RVs, therefore, 12 are required OPERABLE at all times.LCO 3.4.3 requires 12 to be OPERABLE and shutdown if one of the 12 required S/RVs is inoperable. As such, the two Specifications are equivalent and this change in presentation is considered administrative. BFN-UNITS 1, 2, 5.3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF VALVES A5 BFN CTS 4.6.0.3 is only applicable to three stage Target Rock S/RVs.Only the two stage Target Rock S/RVs are'nstalled and authorized for use in BFN Unit 2.The three stage design is obsolete and is no longer supported at BFN.Since this Surveillance Requirement is no longer applicable to the BFN S/RV design, the deletion of this requirement is considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE Ml Ah'additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because't stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.6.D.1).CTS requires a shutdown to MODE 4 within 24 hours but does not stipulate how quickly MODE 3 must be reached.Reference Comment L2 which addresses the less restrictive change of be in MODE 4 in 36 hours rather than 24 hours.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl The details relating to methods of performing Surveillances have been relocated to the Bases or procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs.LA2 This Surveillance Requirement has been relocated to plant procedures since the requirement does not directly relate to S/RV operability. This is strictly a preventive maintenance requirement. 0 QF Pr."-BFN-UNITS I, 2, 5 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF YALVES"Specific" Ll The allowed lift setpoint tolerance has been increased from 1%to 3%based on incorporation of this larger setpoint tolerance in the BFN reload licensing analysis for each Unit prior to ISTS implementation. The larger setpoint tolerance has already been incorporated into the Unit 2 reload analysis and will be incorporated into the Unit 3 reload analysis for the next cycle (Spring 1997).In addition, when the setpoints are verified, they are still required to be reset to 1%(proposed SR 3.4.3.1).Thus, since the analysis still ensure that all limits are maintained even with the expanded tolerance, this change is considered acceptable. This change is also consistent with the BWR Standard Technical Specifications, NUREG 1433.L2 The time to reach NODE 4 (reactor depressurized to<105 psig, Cold Shutdown)has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in NODE 3 (Hot Shutdown)within 12 hours has been added (Reference Comment M4 above).These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.0 BFN-UNITS 1, 2, L 3 Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 (gi SPec ikcrc]ion (g 1.a.3 pp)>crab;I gg~3.R.9.k Mo 3,4,q.c.es Lz w3 Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 e into the primary containment irom unidentified sources shall not exceed In add tion, t e total reactor coolant system leakage into the primary containment shall not exceed sea.v.e l l.Reactor coolant system leakage shall be checked t e ai s li Jal 8 r ore t least once per~hours.b.Qo 3,Q,Q,Q Qi'+in~p JlCvia~haytime the reactor is in RUlf NODE, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged axe~ay 24-hour period in which the'eactor ia in the RUE NODS cep as e ne n 3.6.C.l.c below.C~LCo Z.g,q g During the first 24 hour in the RUE NODE following STAEHJP, an increase in reactor coolant leakage into the primary containment of.,>2 gpa is acceptable as long as the requirements of 3.6.C.l.a are met.Rl kg gmggq~BFH Unit 1 3.6/4.6-9 ANENOMNr N0.I g 7 i~AGE~~>>-~ 8PECi K<C$og Q(QEC 07 19S4 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the-succeeding 24 hours for the sump system or 72 hours for the air sampling system.2 With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.~<<3'us4g;~hon P,~+~~BP'r4 1575 Z,q,q The air sampling system may be removed from service for a L3 period of 4 hours for calibration, function testing, and maintenance without Ad~R+o~-providing a temporary monitor.'A+gCQHl<g~~~8 3.If the condition'n 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed'n~"~4'+W th COLD SHUTDOWN CONDITION within hours.3.6.D hRlig Q 1.When more than onc relief valve is known to bc failed, an orderly shutdown shall bc initiated and thc reactor depressurized to less than 105 psig within 24 hours.The relief~alves are not required to be OPERABLE in the COLD SHPTDO~.ONDITIOH.+ca fD 8I-iv R(g~e pl d JF<guircd'can a,g Add P" once'A'on oP ml Rch'on L 4.6.D l.Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.2.In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is loving from the valve.BFN Unit 1 3.6/4.6-10 AMENDMENT NL R l

UNIT 2 CURRENT TECHNICAL SPECIFICATION IVIARKUP PAGE OF~ ~~la'g j ar v~'.rmzv rv't'tv'A'~~~~s var vr@@sit~ava~'v'~~v:si~~~H.'l0 gv)gl aa IA~~'%'If'LW le (a~~~~~~~~~a~~~~~~~~~I~~~~,~~AI~'I e&~.a~.'\~~~~~~~~~~~~~~~~~~~~~II~~~~I'~~~~~~~~~~~\A~~~~AI~'~~~~~~~~~~~~I I~~~~4P l~I I'll,~~~~~'S I'~~~~I~~~~I't~~AI S~'4 C~4;o~S.V.9 DEC 0'?199'I 2~Anytime irradiated fuel is in the reactor vessel and reactor coolant teaperature is above 212 F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to bc inoperable for any reason, the reactor may reasfn in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.2.Vith the air sampling system inoperable, grab samples shall be.obtained and analyzed at least once every 24 hours.S~>~S 7.]Rmnoe Fog~""~~S~ar Is~~.yS<~IS S'~clod The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout rovidi a tea ora monitor.L3 Adct.Dc7'Iod P, P jeep~'~ego~3.If the condition in 1 or 2 above cannot be metD L11 ,~<<~c.orderly shutdovn shall be]V(d eepu,~&A 4o~B.Z AdcP 2M D Ac7104 Q)(Igf Co<JiSr~3.6.D initiated and the reactor shall be placed i the COLD WK CONDITION vithfn ours~Po7 5//~why 4,oD3plZ3o& JQ gong and When more than one rcl e valve is knovn to be failed, an orderly shutdovn shall bc initiated and the reactor depressurfred to less than 105 psig vf@EX~4 hours The relief valves are not required to be OPERABLE in the COLD SHUTDOW COHDITIOlf ~>3~S7 I/iCA77ohJ p'o~CAAo1665 7o g/Q I~7~g qg~N<<<~5'Cc77og 4.6.D 1.Appro~tely one-half of all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating, cycle.All 13 valves vill have been checked or replaced upon the completion of every second cycle.In accordance vith Speci f ication 1.0.lS, each relief valve shall be manually opened until theraocouples and acoustic monitors dovnstream of the valve fndfcate steam is floving from the valve.BFH Unit 2 3.6/4.6-10 hMENDMBIT N.2 29:=A!."." 3

UNIT 3 CURRENT TECHNICAI SPECIFICATION MARKUP PAGE OF SA c.i<i 0'o AUG 26 1987 l.a.R ppl ifct fy,'tip/CO 3 Lco gg g.~~<2+3 Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 F e c r coo ant ge into thc primary containment from unidentified sources shall not 5 In addition, the total reactor coolant system leakage into the primary containment shall not exceed SR 3,~.w.t 1.Reactor coolant system lcakagc shall bc checked y t sum an ai sam li st re ord at east once per hours.b.y)s&>n%AC FfC Js04LS Anytime the reactor is in RUB mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm avcragcd in vhich the reactor is in the RUE mode cpt as c c 3.6.C.l.c below.Co During the first 24 hour in'he RUN mode follov STARTllP, an increase in reactor coolant leakage into thc primary containment of>2 gye is acceptable as long..as the requirements of 3$%1.a"are met.Pl)Rdd 4C0 3.q,'I,a BPS Unit 3 3.6/4.6-9 NatmNr gO.g o 8 PAGE 3 DEC 0 7'1994 20 R>D/I@ion A+Requircst Avion 8,1 Ac<i'aN C (t s4 Co~d4'i~)t 3.6.D.Anytime irradiated fuel is in thc reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systens shall be OPERABLE.From and after the date that onc of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for.the air sampling system.The air sampliag system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout rovidiag a temporary monitor.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in thc COLD S CONDITIO vithin ours~34 l SNYPo~4 fnndifjoee Ift/P honte md When more than re c valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurised to less than 105 psig within 24" hours.The relimf snLLyes are not required to be OPERABLE in the COLD SHUT1XNN CONDITION. 2.With the air sampliag system inoperable, grab samples shall bc obtained and analyscd at least once every 24 hours.5SLeSAF'eQon &4 Qhagu+8 P'nl Z.S TS Z.q,5~+hiSSccgon D kQLhiF<Qt'.tl4~Bo2"~~><~Qnn(i'h'on oF~)Rch'on 4.6 D.1.hpproxijnately one-halt of all relief valves shall be bench-checked or replaced with a benc~hecked valve each operating cycle.hll 13 valves vill hav been checked or replaced upon the completion of every secoad cycle.BFN unit 3 F44 ipse'cc4Ãon A~~+pcs W Bkd%575 3.9,'3 xn W>s',ekion 3.6/4.6-10 2.Ia accordaace vith Specificatioa 1.0.IR, each relief valve shall be manually opened until thermocouples aad acoustic monitors downstream of the valve indicate steam is flowing from the valve NENMBlT NL I 86 q IP

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE ADMINISTRATIVE A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The total LEAKAGE limit applies at any moment, to the previous 24 hours (not any future or past 24 hour period).This results in a"rolling average" covering"any.24-hour period." Therefore, changing"any" to"the previous" does not change any intent.In addition, the current provision (CTS 3.6.C.l.c), which allows an increase in reactor coolant leakage into the primary containment of)2 gpm during the first 24 hours in the RUN mode following STARTUP as long as unidentified leakage and total leakage limits are not exceeded, is encompassed by proposed LCO 3.4.4.d which allows the same.LCO 3.4.4.d is worded differently (i.e., a 2 gpm increase in unidentified leakage within the previous 24 hour period in MODE 1)but means the same.Since there is no"previous" 24 hour period until being in MODE 1 for 24 hours, this limit does not apply for the first 24 hours.These are editorial changes only and as such are considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE A new requirement has been added to preclude pressure boundary LEAKAGE.An applicable ACTION has also been added.This is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE CTS 3.6.C.3 requires an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION within 24 hours when certain conditions can not be met.Proposed Action C will require the plant be in MODE 3 in 12 hours and MODE 4 in 36 hours.The addition of this intermediate step to the COLD SHUTDOWN CONDITION is considerqd more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.M3 The proposed applicability of MODES 1, 2 and 3 is more restrictive than CTS 3.6.C.l.a applicability of"Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F." The Startup Mode will now include the mode switch position of"Refuel" when the head bolts are fully tensioned. The change eliminates the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist.Currently, CTS 1.0.H allows the plant to be considered in the SHUTDOWN CONDITION and in the Shutdown Mode with the mode switch in the Refuel position (and other positions are allowed while in the Shutdown Mode)as permitted by notes to that definition. The allowance to place the Mode Switch in other positions has been moved to Section 3.10, Special Operations and Section 3.3.2.1, Control Rod Block Instrumentation. Any technical changes to these allowances will be discussed in the Justification for Changes to these Sections.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Details of the methods for performing this Surveillance are relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs.PAGE~Or~BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE"Specific" Ll The total LEAKAGE allowed has been increased to 30 gpm.No applicable safety analysis assumes the total LEAKAGE limit.The limit considers RCS inventory makeup and drywell floor drain capacity.The new limit of 30 gpm is well within the capacity of the Control Rod Drive System pump and the RCIC System, and is well below the capacity of one drywell equipment drain or floor drain pump, which is used to pump the water out of the collecting sump.The collecting sumps can also accommodate this small additional leakage rate.L2 The Frequency has been changed from 4 hours to 12 hours, consistent with the allowance in Generic Letter 88-01, Supplement 1.The supplement allows the Frequency to be extended to shiftly, not to exceed 12 hours.Browns Ferry Technical Specifications currently define the frequency of shiftly as 12 hours, thus, this Frequency is adjusted to coincide with this.CTS do not provide a period of time to reduce leakage prior to initiating an orderly shutdown.Proposed ACTIONS A and B allow 4 hours to reduce LEAKAGE within limits prior to initiating a shutdown.This is reasonable since the total leakage limits are conservatively below the LEAKAGE that would constitute a critical crack size.The 4 hour completion time for ACTION B is reasonable to properly verify the source.of unidentified leakage before the reactor must be shutdown without unduly jeopardizing plant safety.The proposed changes are consistent with the BWR/4 Standard Technical Specifications, NUREG 1433.L4 The time allowed to shutdown the plant when the required actions are not met has been changed from"in the COLD SHUTDOWN CONDITION within 24 hours" to in MODE 3 (Hot Shutdown)in 12 hours and MODE 4 (Cold Shutdown)within 36 hours.The proposed 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.The additional 12 hours allowed to reach Mode 4 is offset by the safety benefit of being subcritical (MODE 3)in a shorter required time.0 BFN-UNITS 1, 2,&3 PAQF~OF~Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE L5 Proposed LCO 3.4.4, RCS Operational Leakage, will add an alternative to existing requirement in Specifications 3.6.C.l and 3.6.C.3 that a reactor shutdown be initiated if unidentified leakage increases at a rate of more than 2 gpm within a 24 hour period.Under proposed Required Action B.2, unidentified leakage that increases at a rate of more than 2 gpm within a 24 hour period will not require initiation of a reactor shutdown if it can be determined within 4 hours that the source of the unidentified leakage is not service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids.This alternative Required Action is acceptable because the low limit on the rate of increase of unidentified leakage was established as a method for early identification of Intergranular Stress Corrosion Cracking (IGSCC)in Type 304 and Type 316 austenitic stainless steel piping.IGSCC produces tight cracks and the small flow increase limit is capable of providing an early warning of such deterioration. Verification that the source of leakage is not Type 304 and Type 316 austenitic stainless steel eliminates IGSCC as a cause of leak.This significantly reduces concerns about crack instability and the rapid failure in the RCS boundary.Also, the unidentified LEAKAGE limit is still being maintained and will continue to limit the maximum unidentified LEAKAGE allowed.This change is consistent with NUREG-1433. BFN-UNITS 1, 2, 5 3 Revision 0

CURRENT TECHNI AL SPECIFICATION MARKUP

~8-4Co 3.9e4 QCT]ddt r90 B Pr'~ca hfoQ W ifcHonS g 8 c dry I~a y 3 Al Anytime irradiated fuel is in the reactor vessel and reactor coolant tempera e is above 12'F, oth the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 urs for the sump system or~~emu's for the air sampling system., e air s pling system ay e r oved fr m serv ce for a per d of 4 hours or cali ation, funct n tes ng, and ma ntenan e vit ut provi a tern ora monit r Zeu w , Rcg'on 8.I$6 QQq5 I I QKC 07 1994 2.With the air sampling system inoperable, grab samples shall be obtained and analyze at least once every~hours./2 3.AhogS C+o If the condition in 1 or Q2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION m tiorA r~~~iT~i<l2howrc one f.e.D vithin h s.3~Ll 1.When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor dcprcssurizcd to less than 105 psig vithin 24 hours.The relief valves are not required to bc OPERABLE in thc COLD SHUTDOWN.~NDITIQN. 4.6.D l.Approximately one-hal of all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating cycle.All 13 valves vill have been checked or replaced upon the completion of every second cycle.See Su~gp;~ye C4lnoIt.5~8F:~bt Bg5 5qcg og 2.In accordance vith Specification 1.0.NM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.BFN Unit 1 3.6/4.6-10 AMENDMNT NO.2 Z3 OF~ S IAblE 3.2.51RtNENIAI ION IIAI INN)ISIS i INIO OR tl 5 stas~2 f I ent 0 qupa ra n F lou Inl ator Smp fill te I iaer Swp Punp Out Rate i Set ints>20.1 ain.I c i ain A~ctl abl se ne r ctor cool I leakage.2.idered par of swp s tea.LAG>"t.S.a Floor Orain F lcm Intcgrator Swy Fill Rale Iiaer Sllql P~1 Out Rate Iiaa:r A.i Ia.8.9 in.I.Used to ghteoaine unidentifiable reactor cool anl leakage.2.Considered part of sunup systea.gLo p q,g b Or@sell Air Sampling L3 c~Oas Pari cu ate backgfoungde I4 C~4 I o NOISES: (I)Qicnever a system is required lo be operable, there shall be one operable systco ellher autaaatic or annual, or7~<T'f>~the acliun required in Section 3.6.C.2 shall be laken.(2)alt ate sy a to de ine the le f lao aanual syst rcby I lee betve swp pwp~tarts is eaka ou because olune o he s u be knae~(3)titan Ageipt of alum, imacdiate acti ulll be lakcn to conf lra lhc ala and assess t~poss y f Increas+leakage. BFN Unit I

Function IIIR IN ES I IABIE 1.2.E%ICY FQt ll tEAII EC N functional Iesl gg q g Calibralion 5 I S R 9.'f 5 J.5.fnstrmi~nchec ~F loor Orain Simp F lm Intciirator Air SuplinII Systea SR 3..5.'K Rale IQ PI~s~once th LS~2 hrS Fino Orain 4'ilg le n l Rath,1laars ~ohqel rail c e oo ra n tag c one qar bicycle y PlORS i,Banal 4 qgly~.7r)blr q,Z.E.~nfl)Ainn hfe65 Are 434 raised i n 4 J)c~tr~f$~gag~3 3~~adjt SR a.e s.z.~zlda~1.Functional tests shall be performed once per DEc, nicAAGA~l.0'J N26 1999 2.Functional tests shall be pcr ormed be ore eac startup with a required frequency not to exceed once per veek.3.This instrumentation is excepted from the functional test definition. The functional test vill consist of in)ecting a simulated electrical si al into th surement channel.4.ested~ing~ogic s tern cti 1 tee~.L,Ay 5.Refer to Table 4.1.B.6.e ic sys func onal tee s shaQ incl+e a cail,ibratkqn ange gcr o erati cycle f time clay rc e anKtimerif ncccs1a fo~propert f th tri s terna The functional test vill consist of verifying continuity across thc inhibit vith a voltohnm)eter. S.Instrument checks shall bc performed in accordance vith the definition of instrument check (sec Section 1.0, Definitions). hn instrument check is not applicablc to a particular setpoiat, such as Upscale, but.is a qualitative check that the instrument ie behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check ie included ia this table for convenience and to indicate that an instrument check vill be performed on the instrument. Instrument checks are not required vhen these instruments are not required to be OPERhBLE or are tripped.9.Calibration frequency shall bc once/year. 10.Deleted 11.Portion of the logic is functionally tested during outage only.12.The detector vill be inserted during each operatiag cycle*and thc proper amount of travel into the core verified.13.Functional test vill consist of applyiag simulated inputs (eee note 3)~Local alarm lights representing upscale and downscale tripe vill be verified, but no rod block vill be produced at this time.The inoperative trip vill be initiated to produce a rod block (SRM and IRM inoperative also bypassed vith the mode evitch in RUN).The functions that cannot be verified to produce a rod block directly vill be verified during the operating cycle.BFH Unit 1 3.2/4.2-59 mml)@sr HS.1 64 PAGE OF 0 UNIT 2 CURRENT TECHNICAL SP ECIFICATION MARKUP p>aa DEC 0 7 89'I poJa J,2t 2~g~l;~l;l,g L.LO i 384 P t'Tlob5 P+Q w.p sQ eA~)4rsoCd 8 hnytime irradiated fuel is in the reactor vessel and reactor coolant tern erature is above 212'F, both the sump and air samp ng systems shall bc OPERABLE.From and after thc a e t at one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for thc'ump system or~muse for the air sampling system.4p~:~2.AC.4-on Sl With the air sampling system inoperable, grab samples shall be obtained and analyze at least once every hours.e air sampling syst may be r oved fro scrvicc fo a per d of 4 h urs for cali ation, f ction test and ma tenance thout providi a tcmpor ry monitor.Ae.Tlsg s c+b.6.D 3~If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall bc initiated and the reactor shall be placed in e COLD WR COKDITIOI vithin hours.3'~+~~TIP ICATJON F'og, CJJhuMc~epnJ>@~~43 rm 7R<J~qy]oQ 1.When morc than onc relief valve is knovn to bc failed, an orderly shutdovn shall be initiated and the reactor dcpressurired to less than 105 psig, vithi~~hours~ The relief valves arc not required to be OPERABLZ in the COLD SHUTDOWNÃCOMITIOS.4.6.D fgo~$pfH79OCdh) Co+I s7lo 12, goal J oat 1.hpproxiaately one-half of all relief valves shal)be bench-checked or replaced vith a bench-checked valve each operating cycle.hll 13 valves vill have been checked or replaced upon the completion of every second cycle.2.In accordance vith Specification 1,O.W, each relief valve shall be manually opened until thcraocouples and acoustic monitors dovnstream of the valve indicate steam is f loving from the valve BFH Unit 2 3.6/4.6-10 AMENMENT NO.Z Z9 PAGE

TASIE 3.2.E INSTRIICNIA IIRT NNIfORS tEAXAGE I ORYKtt System 2 qu pment Ora n Fl ntegrator Swp ll Rate TIa>>r Swp.Pmp t Rate Tie>>r Set ints A>20.1 min.<13.1 min Action I.se o erm ne dent I f e r tor coolant leak@a.gAs 2.Cons red part of swpgystea. F loor Orain Floe Integrator Smp Fill Rate 1 Ia>>r Smp Pmp Out Rate Tie>>r>80.i in%.9 sin.I.Used to determine unidentifiable reactor coolant leakage.2.Considered part of swp system.L<0 Or@el I Air Sanpl ing LE 3 Xgverage background or Gas and Part culate W hl lu I Cl NOTES: (I)lkenever a system'Is required to be operable, there shall be one operable system either autanatlc or manual, I rT><<or the action required In Section 3.6.C.2 shall be taken.A 2)An alte te system to determine the leakage f1~~a manual system<>her~the tie>>betuee~wp pwp tarts ls amitore~The tie>>interval<i&i determine the Ieakag~ou because the~m>>of s vill be knam.{3)Upo~ecelpt of alarm,<<n>>te action uill be taken to conf I@a the alarm and assess the Ibilit f increaL~leakage. BFNMlt 2 Function~TABIE i.2.~ININll'I@1 ANO CAI IBNAI ION FNE NCV tN ORAKLL LEAK OEIECIIQI INSIN~IAI ION sg 3.4s'.)Floor Oraln Sup Flee Integrator Air Sanpllng Systea SR 3..5.>te 3)A5 once ths eaealday L~(g.~LA4 I~a n S te and ce/r c~Ib9L fiber Oralnc'g.Aq I ance/operant cycle 8FNZJnlt 2

W'rg/4 aU<o qpc/-A'~4 0'2 g 7A r~m,~>ivy/a am<~raSs4,>>/4~~~~~yC<kCA~ZQ~)~~~~a/5$3.45.z.Fir~l.Functional tests shall be performed once per P~<<l ical(o'e 3 V.S'JAN 26 1999 20 Functional tests sha e per orme e ore eac startup with a require frequency not to exceed once per week.3~4~This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measur ed during logic stem functiiial teh4s.5.Refer to Table 4.1.B og c system unc onal te~ts shall include a o~ibration~nce ~er ope ing cycle o~e delay relays and timers necessary for proper.functio ng of the tri s stems 7.8.The functional test will consist of verifying continuity across the inhibit with a volt-ohmmeter. Instrument checks shall be performed in accordance with the definition of instrument check (see Section 1.0, Definitions). An instrument check is not applicable to a particular setpoint, such as Upscale, but is a qualitative check that the instrument is behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check is included in'this table for convenience and to indicate that an instrument check will be performed on the instrument. Instrument checks are not required when these instruments are not required to be OPERABLE or are tripped.9.10.Calibration frequency shall be once/year. Deleted Portion of the logic is functionally tested during outage only.12.The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.13.Functional test will consist of applying simulated inputs (see note 3).Local alarm lights representing upscale and downscale trips will be verified, but no rod block will be produced at this time.The inoperative trip will be initiated to produce a rod block (SRM and IRK inoperative also bypassed with the mode switch in RUN).The function that cannot be verified to produce a rod block directly will be verified during the operating cycle.~<<~sf'Pcr4i< 4r FA~qz~~8/575'3, BFN Unit 2 3.2/4.2-59 AMENDMECRV. y jy rAGF~~F~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF~ DEC 07 1994/7lpgg5,\2$3 2~t~Xg.4$(.f LOA 5 P~6 Pcc posed blok K Ac.A 5+8 Anytime irradiated fuel is in the reactor vessel and reactor coolant tern erat re is above 212'F oth the sump an air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or i~oars'or the air sampling system.Req~'nd@thon 8.I 3odAyp 2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every~hours., 12 Th aa.sam ng sys em ma be rem ved from service or a peri of 4 h s for calib tion, f ction te ting and mai tenance ithout rovidin a tern o moni or.3~/}cTlo gg c+u.6 D.BEN Unit 3 If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in he COLD SKJTDOMN CONDITION within ours e Ll When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurised to less than 105 psig within 24 hours.The relief'~ves are not required to be OPERABLE in the COLD SE/TDOMN CONDITION><>>'@AH'o g<Hhsb aP T~ggA/lSTS+hi5 Se~bo 3.6/4.6-10 ,~c Abr SlivgeWe CoAoi5ou>~l2.hours and.6.D.1.Approximately one-hal of all relief valves shall be bench-checked or replaced with a bench~hecked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.2.In accordance with Specification 1.0.NM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valvr NeoMENNO.Z 86 i AcE~oF~' 0 'ThNE 3.2.E TI5TRNKNT THAT ICNTTNS LfhlNiif TIITO tl~iov~aor~4r Rap Out Rate llmr gO.I ale.~i<<T3.4 ale.Used to dateaine Identifiable reactor ant Toalraoe.2.Considered of susp systea.Floor Oraln Flee lnte9rator LCO Susp Fill ILate Tlmr~'"'~I S~FLep Out gaQ Tieer LC.>3i'I.>ibOryuel 1 Alr SapllnII 0'Oa tart cu ate l.Used to date>aine unidentifiable reacior coolant leakage.2.Considered part of susp systea.RIG,: (1)lt>>never a systea ls rawlred to be terable, Mere sl>>II be one cperable systea eitber autcaatlc or aanual, or the action required le fectlon 3.C.CYshall be tahse.{2)An 1 terna ystea deterai the Teakyg~as ls a 1 s a+i ed.T Im Ia al uil detaealee 1 Tat 0>>mba susp s ts ls~J vo suep 1 be{3~rece of al Ieaedla tlon 1 be to coe the a aed sess poss lllty of rea sad e late A Q b

N I 1 Chil Yl F hNE.2.f CB-.floor Orala Sap fly'y4eyrator hir Qep I lac Systea 5g 3.,g.g I2)f5.liS Kqu t la F ala t c KN-lhlt 8 8)CL)CÃl

JAN 26 tS89 Only Noes l,g~d 0 aPP/9++$lbk q.2.E~r<~',.SPeC'afio~ 9..5~o<S~adcb<SccL i n~~r/MALS 6 San+ion p,3~s~tg+55y. 9J Sg, 3ego 5eX day l.Functional tests shall be performed once per 5 Functional tes s e 1 be performed before each startup with a require frequency not to exceed once per veek.3.Thiy instrumentation ie excepted from the functional test definition. The functional test vill consist of injecting a simulated electrical signal into the measurement channel 4.Test dur g lo 5.Refer to Table 4.1.B.al este.J R~J 6.e logic ystem functi 1 tests ha c u e ca b tion~ce ger op ating c le o time de y rela and t ers n cesar for prier)func oaing o he t e st 7.The functional teat vill consist of verifying continuity across thc inhibit wi,th a volt ohmmeter.8.Instrument checks shall be performed in accordance with the definition of instrument check (ece Section 1.0, Definitions). An instrument check is not applicable to a pareicular setpoint, such ae Upscale, but is a qualitative check that the instrument is behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check is included in this table for convenience and to indicate that an instrument check will be performed on the instrument. Instrument checks arc not required when these instruments are not required to bc operable or are tripped.9.Calibration frequency shall be once/year. 10.(DELETED)11.Portion of the logic is functionally tested during outage only.12.Thc detector vill bc inserted during each operating cycle and thc proper amount of travel into the core verified.C 13.Functional test vill consist of applying simulated inputs (see note 3).Local alarm lights representing upscale and downscale trips will bc verified, but no rod block vill be produced at this time.The inoperative trip vill be initiated to produce a rod block (SRM and IRM inoperative also bypassed with thc mode svitch in RUN).The functions that cannot be verified to produce a rod block directly vill be verified during the operating cycle.See Pug>'iak'on &r<tony S Gpss+5+BFN Unit 3 3'/4.2-58 AMENOMERF WP.y P g pAGE~o"~ 0 t JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS LEAKAGE DETECTION INSTRUMENTATION ADMINISTRATIVE Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BMR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 The revised presentation of actions is proposed to explicitly identify that LCO 3;0.3 is required to be entered if all.required RCS leakage monitoring systems are inoperable. This action is consistent with the current requirements and is considered a presentation preference. Therefore, this change is considered administrative. A3 The Table format is being deleted.This change is considered a presentation prefer ence.Therefore, this change is considered administrative. A4 Proposed ACTION B is modified by a note that explicitly states that the provisions of 3.0.4 are not applicable. This explicitly allows a mode change when both the particulate and gaseous primary containment monitoring channels are inoperable. This allowance is provided because, in this Condition, the drywell sump monitoring system will be available to monitor RCS leakage and the compensatory actions for the inoperable system will provide additional indication of RCS leakage.This is an administrative change since existing Technical Specifications do not have an explicit requirement that prohibits entry into a Mode or condition when an LCO required by that Mode or condition is not satisfied. Therefore, CTS allows the actions being permitted by the note being added.This is consistent with NUREG-1433, BFN-UNITS 1, 2, 5 3 Revision 0 A5 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION Frequency has been editorially changed from monthly to every 31 days and from every six months to every 184 days.This is an administrative change since these are equivalent time periods.A6 The current provision (CTS 3.6.C.2, 2nd paragraph) that allows the air sampling system to be removed from service for a period of 4 hours for calibration, functional testing, and maintenance without proyiding a temporary monitor has been eliminated. There is currently no requirement for a monitor for at least 24 hours (CTS 4.6.C.2).Therefore, the current provision serves no purpose.TECHNICAL CHANGE-NORE RESTRICTIVE The proposed applicability of NODES 1, 2 and 3 is more restrictive than CTS 3.6.C.l.a applicability of"Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F." The Startup Node will now include the mode switch position of"Refuel" when the head bolts are fully tensioned. The change eliminates the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist.Currently, CTS 1.0.H allows the plant to be considered in the SHUTDOWN CONDITION and in the Shutdown Mode with the mode switch in the Refuel position (and other positions are allowed while in the Shutdown Mode)as permitted by notes to that definition. The allowance to place the Mode Switch in other positions has been moved to Section 3.10, Special Operations and Section 3.3.2.1, Control Rod Block Instrumentation. Any technical changes to these allowances will be discussed in the Justification for Changes to these Sections.M2 The frequency of grab sampling with the air sampling system inoperable has been increased from 24 hours to 12 hours.A grab sample once/12 hours provides adequate information to detect leakage during the extended (See Justification for Change L4)period of time that the air sampling system is allowed to be inoperable. H3 Not used.M4 Not used.H5 The Frequency of the channel check requirement has been changed from every 24 hours to every 12 hours, consistent with Generic Letter 88-01, Supplement 1 and NUREG-1433. This is an additional restriction on plant'peration. BFN-UNITS 1, 2, 8L 3 Revision 0 ' JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION CTS 3.6.C.3 requires an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION within 24 hours when certain conditions can not be met.Proposed Action C will require the plant be in MODE 3 in 12 hours and MODE 4 in 36 hours.The addition of this intermediate step to the COLD SHUTDOWN CONDITION is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating expe}ience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LA1 The description of an acceptable alternate system to measure leakage has been relocated to the Bases or procedures that support compliance with the limits for RCS Operational Leakage in proposed Specification 3.4.4.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures and FSAR will be controlled by the provisions of 10 CFR 50.59.LA2 The details relating to the setpoints have been relocated to the procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA3 The details relating to actions required upon receipt of an alarm have been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA4 Details of the specifics of the functional, calibration, and logic system functional test related to the floor drain sump fill rate and pump out timers has been relocated to procedures since the operability of the system is not dependent upon these timers.Changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 5.3 PAQE 0F Revision 0

LA5 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION The drywell equipment drain sump monitoring system functions to quantify identified leakage.Since the purpose of this specification is to provide early indication of unidentified RCS leakage, the drywell equipment drain sump monitoring system has been relocated to the Bases or procedures that support compliance with the limits for RCS Operational Leakage in proposed Specification 3.4.4.The design features and system operation are also described in'the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures and FSAR will be controlled by the provisions of 10CFR50.59. TECHNICAL CHANGE-LESS RESTRICTIVE"Specific" Ll L2 The time allowed to shutdown the plant when the required actions are not met has been changed from"in the COLD SHUTDOWN CONDITION within 24 hours" to in MODE 3 (Hot Shutdown)in 12 hours and MODE 4 (Cold Shutdown)within 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added.These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.This requirement has been deleted.An instrument check would not consistently demonstrate operability since normally the instruments could not be compared to any other instruments, and their reading could be anywhere on scale;thus, observing the meter would provide no valid information as to whether the instrument is OPERABLE.The CHANNEL FUNCTIONAL TEST requirement is the best indicator of OPERABILITY while operating, and this requirement is being maintained. This is also consistent with the BWR Standard Technical Specification, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION L3 CTS Table 3.2.E defines the air sampling system as consisting of gas and particulate monitoring channels (i.e., both channels are required OPERABLE for the air sampling system to be considered OPERABLE). .Proposed LCO 3.4.5.b requires either one channel of the gas or one channel of the particulate monitoring system to be OPERABLE.This is less restrictive than CTS requirements but is acceptable since either channel is capable of indicating increased LEAKAGE rates thaf correlate to radioactivity levels of 3 times average background. L4 The allowed outage time for the air sampling system has been changed from 72 hours to 30 days.The 30 day allowed outage time recognizes that at least one other form of leak detection is available (sump monitoring) and takes credit for the increased sampling frequency of 12 hours (versus CTS of 24 hrs).This change is consistent with NUREG-1433.L5 The calibration frequency has been changed once pe}3 months to once per 18 months.This new Frequency is consistent with BFN setpoint methodology, which considers the magnitude of the equipment drift in the setpoint analysis over an 18 month calibration interval.The primary containment leak detection noble gas and particulate monitor is a digital Eberline continuous air monitor (CAM)which is identical to the building effluent monitors whose calibration frequency is 18 months in accordance with the Offsite Dose Calculation Manual (ODCN)and previously required by Technical Specification Table 4.2.K until these instruments were removed by Amendment No.216 dated September 22, 1993 (reference TS 301).Excessive calibration can cause damage to the equipment. In addition, plant operations could be impacted while the equipment is removed from service for calibration since it would not be available for leak detection. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE S GF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP clPi'z.V.4 J N2819%3.6.8.4~When the reactor is not pressurized vith fuel in the reactor vessel, except during the STARHJP COHDITIOR, the reactor vater shall be maintained vithin the folloving liaita.4.6.B.4~Whenever the reactor is not pressurized vith fuel in the reactor vessel, a ,saaple of the reactor coolant shall be analyzed at least every 96 hours for conductivity, chloride ion content and pH.ao Conductivity-10 yeho/ca at 25 C b.Chloride-0'ppa c.pH shall be betveen 5.3 and 8.6.Sec Sufkif eeHon Ar CJ4nst 5~'<<><8/9<8.ta this Sec gioa llnK A~sR3'.v,q,l 5~LCO 6.P 9.4 Ryl'c When the tiae liaita or maxiinm conductivity or chloride concentration liaita are exceeded, an orderly ahutdovn shall be initiated iaaediately. The reactor shall be brought to the COLD SHUTDOWN CORDITIOR aa rapidly aa cooldovn rate eraita.Wheneve the reactor ia r t cal the ta on activity concentrations in the reactor coolant shall not exceed the equilibritm value of 3.2 pCi/ga of dose equivalent I-131.CQuirc During guilibriaa pover operati an iaotop c ana aia nc y q t t tive ur cata for at 1 aat I 31-132 I-I-13 be perfoza eel~on a coolan iqaid aaaple.da)s A.+s.l k4ditional coolant samples shall be taken vhenever the reactor ctivi ceeda pre to equilibritm concentration specified in 3.6.B.6 c ti re t: BFR Qnit 1 3~6/4~6-7 AMENOMENT Ra.2 0 8 8~' ympmg N A.+ikey<<,'~s 4r Cu d.h'hA pcs(od@Tfog'8 This limit ma bc exceeded for a maximum of 4$hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gm encver e reac or s cr t cal s pcr tc mor 5X its earl po er op rati n er s cepti n fo thc li activi limits If the iodine concentratioa in the coolant exceeds 26 pCi/gm, the reactor shall be shut dovn, and the steam line isolation valves LQ, ufik4n tQ Qurg or be fn NM'I i0hi~34 h~rZ~f a.During the ST C HDITIOI b.ollovi a si fican over e40 c.,F lloving an,incre c in the equ librium of gas le 1 exceed ng 10, 0 pCi/ec (at the ste get ai e)ector)vithin a 48-hour period d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be takea at 4 hour interval or our, or ti a sta le iodine oncen rati a be ov the limit valu (3.pC ga)i establ shed.Hov cr, a least cons cutive samples shall b akea in al cases.kn isotopic analysis 11 be performed for each sample, and quantitative measuremcnts made to determine the dose equivalent I-131 concentratioa. 7.When there ia no fuel ia the reactor vessel, technical specification reactor coolant chemistry limits do not apply.7.When there is no fuel in the reactor vessel, sampling of reactor coolant chemistry at technical specification frequency is not required.**or the p rpose o this ection sampl frequ cy, a s ficant over chang>>is de ed as a change ceasel 15X f rated p ver ia ess t 1 hour.BFI Unit 1 3.6/4.6-8 NENOMENT lE 2 9 8 PAGE 3 QF~ 0 INSERT PROPOSED NEW SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. PAGE~OF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.4.7 RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN ECHNIC L C GE-0 ST IC IV Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 H(INSERT PROPOSED NEW SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications.

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTBI-COLD SHUTDOWN TECHNICAL CHANGE-ORE ES CTIV Hl A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent. with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS I, 2, 5 3 Revision 0 I 0 UNIT2 CURRENT TECHNICAL SPECIFICATION MARKUP JUNP,8 tsar 3.6.B.Coo 4.6.B.Coo e t@CO 34@hppl c,.4.When the reactor is not pressurized vith fuel in the reactor vessel, except during the STARTUP COHDITIOH, thc reactor vater shall be maintained vithin the folloving limits.a.Conductivity-10 pmho/cm at, 25 C b.Chloride-0.5 ppm c.pH shall bc betveen 5.3 and 8.6.5.When thc time limits or maxilla conductivity or chloride concentration limits are exceeded, an orderly shutdovn shall be initiated immediately. The reactor shall be brought to thc COLD SHUTDOMH COHDITIOH aa rapidly as cooldovn rate crmits.~t 6.Whenever thc reactor i rit ca thc limits on activity concentrations in the reactor coolant shall not exceed thc equilibrium value of 3,2 pCi/gm of dose equivalent I-131.4.Whenever the reactor is not pressurized vith fuel in thc reactor vessel, a sample of the reactor coolant shall bc analyzed at le'ast.every 96 hours for conductivity, chlori ion content and pH.5ec'X<sf<gi'c bio~+<+~d~far<TS 3.C.8/O'.C.g His S~4<o~4r$~3 H.9 I sR 3.'A6 I 5.Dur quilibrium pover o cratio an iaotop c analysis c itativc me urgents for at cast I-13K, I-132-133 and I-1 s be performed on a coolan liquid sample.H3 9 Joys LI j~.+'5, 6.Add tional coolant samples ahall be taken vhcnever the reactor activity exceeds cent equilibrium concentration 4.Al specified in'3.6.B.6 one~f o o c iti are+t: PAGE BFH Unit 2 3.6/4+6-7 NENOMENT RtL Z 2g o 0 (Q S cc,S;,4)..~ ~.q(JUN 8 8$994 p Cl'lo fJ p ETIO+0 Pygmy>gQ/vog~P kfNet'4J Ac~gQ,h c This limit may be exceeded 0 for L.of 48 hours.During this Lctivity transient the iodine concentrations shall not exceed 26 pCi/gm enever t e reactor is critical.e eac operated more t 5X of s year pover eration er s excep ion for e e uil brium ctivit limits If the iodine concentration in the coolant.exceeds 26 pCi/gm, the reactor Shall be shut dona, and the steam line isolation vLlves c W.~(2.w~goal Pep, A.t a.During the STARTUP COHDITIOH Folloving i4fgnffic pover change+c.olloving an in ease i the equilibri off as level excee ing 10,0 pCi/sec (at the steam et Lir e)ector)vi in a 48-hour eriod.d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hour intervals for 48 ours, or unt 1 o~be I'~AW Q~i&5$4v~a sta e iodine c entration belo the limiting va e (357, yCi/gm is established. Hovever, least 3 consecutive samples sha 1 cases ka isotopic analysis Shall be performed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concentration. Shen there is no fuel in the reactor vessel, technical specification reactor coolant chemistry lild.ts do not apply.7.When there is no fuel in the reactor vessel, sampling of reactor coolant chemistry at technical specification frequency is not required.For the purpose of this section sampling frequ, a si ficant poser e i defin as L change ceding 1SX of r ed pover in less than 1 hour.3'/4.6-8 AMEMOMENt'Mt. M M M PAGE

INSERT PROPOSED NEM SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. PAGE OF JUSTIFICATiON FOR CHANGES BFN ISTS 3.4.7 RHR SHUTDOWN COOLING SYSTEN-HOT SHUTDOWN T C NIC L C GE-0 EST IC IV Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BMR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0

INSERT PROPOSED NEM SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Speci fi cati ons.

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTEN-COLD SHUTDOWN TECHNICAL CHANGE-NORE RESTRIC IVE Hl A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2,&3 Revision 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP

3.6.B.4.6.B.C o 4.Mxen thc reactor is not pressurized vith fuel in the reactor vcsscl, except during the SThRTUP COHDITIOS, thc reactor vater shall be maintained vithin the folloving limits.a CoILductivity 10 pmho/cm at 25~C b.Chloride-0.5 ppa c.pH shall be betveen 5.3 and 8.6.4.Whenever thc reactor is not pressurized vith fuel in thc reactor vessel, a sample of the reactor coolant shall bc analyzed at least every 96'hours for conductivity, chloride ion content and pH.~><5+i~i~on Qe Changes'CI'~F 4 8/y.C..8;< +h;s Sccdon Ilfok r 5 3 Q.g, I 5.When the time limits or mm~mL conductivity or chloride concentration limits are exceeded, an orderly shutdovn shall be initiated immediately. The reactor shall bc brought to the COLD SHUTDOWN CONDITION as rapidly as cooldovn rate permits.(6.Whenever the I ritic the limits on activity 3,q,p concentrations in the reactor , coolant shall not cxcecd the (hyllcrk'IkPOoflibrf a ooloe of 3.2 PCi/ga of dose equivalent I-131.5~6~During qu libriua povc peration an sotopic aILa s s 2 tit tive eas cm ts fo at 1 ast-131 I-32, I-1 I-1 4 s bc performed oIL a coolant 1 quid sample.7$/5 e'.Ron,l 8.1 0 coo ant samples shall be taken vhenever the reactor activity exceeds 0 e en 0 e equ br ua concentration specified in 3 o th fo ovi c i i a m BFR Unit 3 3'/4'-7 AMENDMENT NL y 8 y F~GF~OF

n3.%6 jets)sg Qhon P (I+fon 8lo gpooS A4 oo gcpw o',CcP j44s'y~S 4w (e J4 A This liuent Iaay bc exceeded for a aaxiam of 48 hours.During this activity transient thc iodine concentrations shall not exceed 26 pCi/eaevcr reactor s cr tical e e ore o rate r than 5 of i s yea y vc opcrat oa er s e ion, for c c ilib a iv ty liaits If the i dine concentra on in the coolant exceeds 26 pCi/ga, thc reactor ahall bc shut down, and the steaa line isolation valves shall be clos bligh'on~ghmr or be on Q~9 lPo'Hlo&3Q~Op AC), At'H n k.l a.Duri the ST CO ITIOH b.Follovi a signi f ant pover ec*C~llovtng an increase the equili ri~of gas level ceding 10, 0 pCi/sec at the ste 5et air C5 tor)vithia a 48-hour riod.d.Whenever the equilibrhm iodine limit specified in 3.6:B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hour intcrv 1 o 48 ho s, r until stab iod coac trc ioa b lov liat v e.2 p/gR)s cata ished Ho ver, t least 3 c cu s ess lbet al cases isotopic analysis be pcrforaed for each saaple, snd quantitative aeasureaents sade to detcraine the dose equivalent I-131 concentration. 7.When there is no fuel in the reactor vessel, technical specification reactor coolant cheaistry limits do not apply.g~>WAX aNon 4 r C~g'>l'i~B (~oo,;s rC~7.When there is no fuel in the reactor vessel, saapling of reactor coolant cheaistry at technical specification frequency is not required.For the urpose f this section ssRpl f rcqu cyg a s fican power is dc cd as a change ccd 15X f rated vcr jn css 1 hour.BFI Unit 3 3.6/4.6-8 AMENOMgg~L Z 81 Gp g 0

INSERT PROPOSED NEW SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. ~~CE OF JUSTIFICATION FOR CHANGES BFN ISTS 3.4.7 RHR SHUTGOMN COOLING SYSTEM-HOT SHUTDOWN EC NICA CHANGE-0 E EST IC I Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS I, 2, 5 3 Revision 0

y(INSERT PROPOSED NEM SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN ECHNICA CHANGE-ORE ESTRIC IVE Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent.with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 OF

13USTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY ADMINISTRATIVE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical, Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.~A2 Note is added to the Required Actions for Condition A to indicate that LCO 3.0.4 is not applicable. Entry into the Applicable Modes should not be restricted since the most likely response to the condition is restoration of compliance within the allowed 48 hours.Further, since the LCO limits assure the dose due to a LOCA would be a small fraction of the 10 CFR 100 limit, operation during the allowed time frame would not represent a significant impact to the health and safety of the public.In addition, this allowance is already inherently provided by the words of Specification 4.6.B.6.a, which states that additional samples are required"during startup" when specific activity exceeds the limit.Thus, this change is a presentation preference only and is considered administrative. A3 Existing Specification 3.6.B.6 requires that if the Dose Equivalent I-131 cannot be restored within 48 hours, or if an any time it exceeds 26 pCi/gm, the reactor must be shut down and all main steam lines must be isolated immediately. Proposed LCO 3.4.6, Condition B, allows the alternative of being in MODE 3 within 12 hours and Mode 4 within 36 hours under the same conditions. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).In Mode 4, the LCO requirements are no longer applicable. This change is considered administrative because existing 1.0.C.1 would require that the reactor be placed in Mode 4 within 36 BFN-UNITS 1, 2, 8L 3 Revision 0~u-;D JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY hours if the requirements in CTS 3.6.B.6 could not be met.This change is consistent with NUREG-1433. TECHNICAL CHANGE-MORE RESTRICTIVE Ml~e The Applicability has been changed to require the specific activity to be within limits in those conditions which represent a potential for release of significant quantities of radioactive coolant to the environment. Thus, MODE 3 with any steam line not isolated has been added.In addition, MODE 2 with any steam line not isolated has been added in lieu of MODE 2 when the reactor is critical.While this does allow the reactor to be critical with the main steam lines isolated while not requiring the LCO to be met, overall this change is considered more restrictive due to the MODE 2 subcritical and MODE 3 requirements. In addition, the ACTIONS have been modified to reflect the new Applicability, and an option for exiting the applicable MODES is-provided for cases where isolation is not desired.CTS 4.6.B.5 requires sampling reactor coolant to determine specific activity"during equilibrium power operation." Proposed SR 3.4.6.1, which contains proposed requirements for sampling reactor coolant to determine specific activity, is modified by a note that requires this Surveillance to be performed only in MODE 1.This change is slightly more restrictive because sampling will be required whenever the reactor is in MODE 1 and not just when equili6rium conditions have been established. This change is consistent with NUREG-1433.- M3 The Surveillance Frequency has been changed from monthly to weekly (every 7 days)for consistency with NUREG-1433, Rev.1.Since Revision 1 to the NUREG deleted the surveillance requirement to verify that reactor coolant gross specific activity is less than or equal to 100/E-bar pCi/gm every 7 days, the reactor coolant specific activity trending interval was decreased to 7 days from 31 days.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" Revision 0 LAl CTS 4.6.B.6 contains requirements for reactor coolant and offgas system~~~sampling during startup, following significant power level changes, and BFN-UNITS 1, 2, 8L 3 2 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY following significant changes in offgas radiation levels.The results of any of these samples are intended to determine if RCS specific activity is exceeding specified limits.Experience has determined that the weekly sampling required by proposed SR 3.4.6.1 and requirements for monitoring main steam line and offgas radiation levels is sufficient to ensure RCS specific activity levels are not exceeded.Therefore, RCS specific activity requirements for sampling stack gas, offgas and main steam line are being relocated to plant procedures and will be controlled in accordance with the licensee controlled programs.In addition, the criteria for when specific activity has been returned to limits (for 48 hours or until a stable iodine concentration below the limit has been established with at least 3 consecutive samples being taken in all cases)has been relocated to plant procedures and will be controlled by the licensee controlled programs.The method of determining dose equivalent I-131 (i.e., quantitative measurements of specific isotopes of Iodine), as described in CTS 4.6.8.5, has also been relocated to plant procedures. These changes are consistent with NUREG-1433.t"Specific" Ll Pro posed ACTION A allows the LCO limit to be exceeded for 48 hours provided that the specific activity does not exceed 26 pCi/gm.CTS 3.6.B.6 allows the limit to be exceeded during a power transient and limits the time the reactor can be operated, when the LCO RCS Specific Activity limit is exceeded, to less than 5%of its yearly power operation. Generic Letter 85-19,"Reporting Requirements on Primary Coolant Iodine Spikes," states that this limit is not necessary because reactor fuel has improved significantly since this requirement was established, and that proper fuel management by licensees and existing reporting requirements for fuel failures will preclude ever approaching this limit.Removal of this limit is consistent with the BWR/4 Standard Technical Specifications, NUREG-1433, requirements. L2 CTS 3.6.B.6 requires the reactor to be shut down and the'team line isolation valves to be closed immediately if the iodine concentration exceeds 26 pCi/gm.Proposed ACTION B allows 12 hours to close the isolation valves or to be in Mode 3.The 12 hour Completion Time is reasonable, based on operating experience, to isolate the main steam isola'tion valves, or to achieve the required plant conditions, in an orderly manner and without challenging.plant systems.The less restrictive 12 hour Completion Time is consistent with NUREG-1433. BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP PAG~ LI HG COHDITIOHS FOR OPERATIOH Rl SAd>Pi'yn 3.SURVEILLAH QUIREMEHTS 3.6 4.6 Applies to the opcrati status of the r actor coolant stem.Applies t the period cxaminatio and test rcquireacnt for the rca tor coolant syst To assure the tegrity and sa operation of the eactor coolan systems To etcrmine the ondition of the eactor coolan system and the o cration of th safety device related to it.Lgy>q,q 1.The average rate of reactor coolant temperature change during normal heatup or cooldowa shall not exceed 100'P/hr when averaged over a one-hour period.1.During heatups and s R i.q.q,)ooldovna the folloving parameters shall be recorded and ,reactor coolant temperature determined a minute intervals ti success rea at each given ocation are vi 5 P.a.Steam Dom Pressure (Convert o upp r vessel r gion temper ure)Reac r bot om drain tern eratur c.circul ion lo s A and B d.React r vcsse bottom head tempera re e.Rc ctor ve el shell a gaccnt shell flange BFH Unit 1 3'/4.6-1 PAGE OF A<2.Daring all operations vith a critical core, other than for lov-level physics tests, except vhen the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve 03 of Figure 3.6-1.2.Reactor vessel metal temperature at thc outside surface of thc bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to shell flange shall bc ecorde at least every minutes daring inservice bydrostptic or leak testing cn e vessel ressure is>312 psig.3.Daring heatup by nonnuclear means, except vhen the vessel is vented or as indicated in 3.6.k.4, daring cooldovn folloving~'"'l naclear shutdovn, or daring lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of carve 0?of Figure 3.6-1 until removing tension on the head stud bolts as specified in 3'.k.5~R2 3.est ecimcns repre cnt the reactor vcss, ba e veld and vel heat ffect xone met 1 be tailed in e r actor esse a scen to ves 1 v at the re mi plane 1 vel.The amber and e spe ens il b in accor ance GE repo t 1011.e spe ens hall ce the int t of ESTD 1-82.;.":-3 BFS Unit 1 3.6/4.6-2 IIIENDlNENT HO.17 0 SEP I 3 1995 4~L~o Z.9Q SR~~"fi1, NoQ Z, 5.sg 8.'f.$.5/V inc'2 The beltlinc region of reactor vessel temperatures during inscrvice hydrostatic )or leak testing shall be at or above the temperatures shovn on curve 41 of Figure 3.6-1.The applicability of this curve to these tests is extended to nonnuclear hcatup and ambient loss cooldovn associated vith these tests only if the heatup and cooldovn rates do not exceed 15 P per hour.Thc reactor vcsscl head bolting studs may bc partially tensioned (four sequences of the seating pass)provided the studs and flange materials are above 70 F.Before oading the flanges any more, thc vcsscl flange and head flange must bc greater than 80 P, and must remain" above 80'P vhilc under full tension.Nl~k, I P~pg~sw 3'.Q.5.2.S 0 z.9 f.g+8 a I<<~.s.9.v+Ivy 5'R 3oH.9 1+No4e 5.When the reactor vessel head bolting studs are tensioned and the reactor is in a cold condition, thc reactor vcsscl eratur inacdiately belov the head flange shall be fiopssrd 4<qgswlcs A r SRs 3.q,9.5)g+g BFS Unit 1 3.6/4.6-3 AMENDMBlT NO.2 P.f PAGE~OF~

S'kc'0'ca on 3.'f.9 6.~n gy,q The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant vithin the idle operati ecirc at, on loo s are vithin 50'F of each other.t SR Z.S8.S+~o<X 6.Prior o r R3 startu o an e recirculation loop, the temperature of the reactor i coolant n the o era and idle loops shal e en ly o e L,Co 7~The reactor recirculation pumps shall not be started unless the coolant temperatures betveen the dome and the bottom head drain are vithin 145 F.7.Prior to starting a+'recirculation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall be compared e o H3 P'.po'd Acti<~34s+;F;wg'<<gg, $r BFA 1575 3.q.~S~aMi'cz4on @r Cj~~ggc 4i BFQ tSvs z 4.6.E.~Jt~g~2.Whenever there is recirculation flov vith~the reactor in thc STARTUP or RUH Mode and one recirculation pump i is operating, the diffuser to lover plenum differential prcssure shall be checked daily and the differential pressure of an individual jet pump.in a looy shall'ot vary from the mean of all Jet pumy differential pressures in that loop by more than 10K.3.6.F 4.6.F The reactor shall not bc operated vith onc recirculation loop out of service for morc than 24 hours.With thc reactor operating, if onc recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWN COHDITIOH vithin 24 hours unless the looy is sooner returned to service.l.Recirculation pump speeds shall be checked and logge at least once pcr day.Sg 3Aegi Q 2~3~Folloving onc pump operation, the discharge valve of the lov speed pump may not be opened unless the speed of the faster pump is less thaa 5OX of its rated syeed.When the reactor is not in thc RUN mode, REACM POWER DPERATIOH vith both recirculation pumps out of-service for to 2 hours is yermitted During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is vithin 75'F of the saturatioa 2.Ho additional surveillance required.se~.~.v.v 3.Before starting cithcr recirculation pump during REACTOR PO OPERA the oo s rge temperature and dome saturation tempcraturc. BFH Unit 1 3.6/4.6-12 AMENDMEHT NP.2 y y pAGE~GP~ 0 3.6.F 3.9~9.'9~A/os 2 temperature of the reactor vessel vater as determined b dome pressure.e total elapsed time aa ural circulation aad oae pump operation must be no greater than 24 hours.S~e Y~swkc~g'on @kc 9t-Iv f 5 pg p,q,~The reactor shall not be operated vith both recirculation yumya out-of-service vhile the reactor'ia in the RUB mode.Folloving a trip of both recirculation pumps vhile in the RUN aode, immediately ate a manual reactor scram.3.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained ia accordance vith Syecificatioa 4.6.G throughout the life of the plant.a Mith ths structural integrity of any ESNE Code Class 1 equivalent coaponent, vhich is part of th>>primary system, not confozILing to the above requirements, restore the structural integrity of the affected component to vithin ita limit or maintain the reactor coolant system in either a Cold Shutdova condition or less than 50'F abide the minimum temperature required by HUT considerations, until each indicatioa of a defect has been investigated and evaluated. 4.6.G I l.Iaserrice inayectioa of MME Code Class 1, Class 2f aad Class 3 components shall be performed ia accordance vith Section XI of the ASIDE Boile aad Pressure Vessel Code aad applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except Mere syecific vritten relic haa baca graated by HRC yursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additioaal inayectiona shall be perforaed on certain circumferential yipe velda to provide additional protection against pipe vhip, vhich could damage auxiliary and control systems.5<<3KHiAczg'q ~*'TS S,S.g/q,g y BFS Unit 1 3.6/4.6-13 AMENOggPNg, p p 6;pA'8 OF~~

ClS gg JUN 2 8 1994 3.6.B.1.PRIOR TO ST TUP and at steaming rates'ess than 0,000 lb/hr, the folloving limits sh 1 apply.4.6.B.C o 1.Reactor coolant shall be continuously monitored for conductivity except vhen there is no fuel in the reactor vessel.a.Cond ctivity, pmh/cm at 25 C 2.0 b.oride, ppm 0.1 Whenever the continuous conductivity monitor is inoperable, a sample f reactor coolant hall be analyze for conduct vity every 4 hour except as liste belov.If the react r is in COLD SHUT OMH COHDITIOH, a s e of reactor co ant shall be lyzed for nductivity every hours.b.Once a veek the continuous monitor shall be checked vith an in-line flo cell.This in-line conductivity calibration hall be performed e ery 24 hours vhen ver the reactor c lant conductiv ty is>1.0 pmho/cm t 25 C.2.At steaming rates greater than 100,000 lb/hr, the folloving limits shall apply.a.Conductivity, pmho/cm at 25 C 1.0 2.During, star p prior to pressurizi the reactor above atm pheric pressure measurements of reac r vater quality shall performed to shov confo ance vith 3.6.B.1 of li iting conditions. b.Chloride, ppm 0.2 BFH Unit 1 3.6/4.6-5 AMENOMENT RO.2 0 8 FASP/OF~~ DEC 0 7 L994 3.6.B.4..B.Coo 3~ht steaming rates greater 100,000 lb/hr, thc reactor vatcr qua ity may cxcecd S cification 3.6.B.2 nly for the time 1 its specified belov.Exceeding these time imits or the follow ng max quality limits s all be use for placing the reactor in the CO SHUTDOMR CO ITIOH.a.Conductivity time abov 1 pmho/at 25'C-2 v ks/year.Maxim Limi t 10 mho/cm at 25'C 3.whenever thc reactor is operating (including HOT STAHDBY CORDITIOH) m suremcnts of reactor ter quality shall.be erformed according to the folloving schedule: a.Ch1oride ion content and'H sha be measured least once every 96 ours.b.Chlori e ion conte t shall be meas rcd at least eve 8 hours.vh ever reactor c ductivity is.0 pmho/cm t 25 C.b.Chlor de co entration time a ove 0.2 ppm-2 weeks/year. Maximum Limit-0~5 ppmo c.The reactor shal be placed in thc S WH COHDITIOR if pH<5.6 or>8.6 for a 24-hour period.c.h sample of actor coolant sha bc measured f pi at least onc every 8 hours vh ever the reactor oolant conduct ity i,s>1.0 pmho/at 25 C.BFH Unit 1 3.6/4.6-6 AMENOMEHT NIL R 1 3

~~.e.c'.e,s'ON28m 3.6.B.4.When t c reactor is not p csaurizcd vith fu in thc eactor vessel, cx pt dur thc SThRHJP CO ITIOH, the reactor vater s 1 be ma ntaincd vithin t f lloving limits.Conductivity 10 pmho/cm t 25'C 4.6.BE 4~cnevcr the eactor not ressurizcd ith fuel in the reactor csacl, sample of e react r coolant s ll bc lyzcd at least every 96 ours for conductivity, chloride ion content, and pH.b.Chloride 0.5 ppm c.pH sha be betvccn 5.3 8.6.5.When th time limits r conductivity or chlor de concentrat on limi s are cxceede , an ord ly shutdovn ll be in iatcd imacdia ely.The re ctor shall be brought to the COLD SHUTDOWNS COHDITIOH aa rapidly aa cooldovn rate ermita.5.During equilibrium pover operation an isotopic analysis, including quantitative miasurcmcnta for at least I-131, I-132, I-133, and I-134 shall be performed monthly on a coolant liquid sample.6.Mxcnever the reactor, is critical, thc limits on activity concentrations in the rcictor'oolant shall not exceed the equilibrium value of 3.2 pCi/gm of dose equivalent I-131.6.Additional coolant samples shall be taken vhencvcr thc reactor activity exceeds one percent of the equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are met: BFS Unit 1 3.6/4.6-7 AMENOMENT gg, p O 8 PAep 3.6.B 4.6 3.6.B.6 (Cont'd)This limit may be exceeded folloving pover transients for a maximum of 48 hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gm vhenever the reactor is critical.The reactor shall not be operated more than 5X of its yearly poser operation under this exception for the equilibrium activity limits.If the iodine concentration in the coolant exceeds 26@CD/gm, the reactor shall be shut down, and the steam line isolation valves 11 be closed immediately. 4.6.B.6 (Cont'd)a.During the SThEEP COHDITIOH b.Folloving a significant ponr change~+c.Folloving an increase in the equilibrium off-gas level exceeding 10,000 pCi/sec (at the steam jet air ejector)vithin a 48-hour period.d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hou intervals for 48 hours, or until a stable iodine concentration helot the liNLiting value (3.2 pCi/ga)is established. Honver, at least 3 consecutive samples shall be taken in all cases.ka isotopic analysis shall be performed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concentrations 7.When ere i no fu he ,re tor v sel, chni 1 s ecifi tion r seto coolant eais ry 1 s do ot apply.7.Wh there no uel reacto vess 1)ling f rei tor olant eNList at t chni 1 specifi tion freq ency is not required.**For the purpose of this section on sampling frequency, a significant pover exchange>>is defined as a change excee~lng 15Z of rated pover in less than 1 hour BFH Unit 1 3.6/4.6-8 NENDMENT R5.Z 0 9 N Cl CTS p.6,g NY3>m 3.6.F 3.6.F.3 (Cont'd)temperature of thc reactor vessel water as determined by dome pressure.The total elapsed time in natural circulation and one pump operation must be no grcatcr than 24 hours.St.'c Wus~Flc<A'a~ g,(g)9 8F~ISTIC 3,1')Rc c'<<culm'ova ~Ps oPt<eh'~)i w+gs sccw>n 4.The reactor shall not be operated with both recirculation pumps outof-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G 4.6.Q 1.The st ctural integri of ASME Code C ss 1, 2, and equivalent compon ts shall be intaincd in acco cc with Spe fication 4.6.G thro out the lif of the plant.a.Mith the stra tural integrity of any ASME ode Class 1 equivalent omponent, which part of th primary system, ot conform to the above requirem ts, restore the structur 1 integrity of e affectc component to w hin its 1 t or maintain e react coolant syst in eithe a Cold Shutdo cond tion or less 50'F abov the minimum t peraturc required by HDT co iderations, until each indication of a defect has been invcstigated-and evaluated. Inservice inspection of AS Code Clas 1, Class 2;and Class 3 omponents shall be perform d in accordance with Sectio XZ of the ASME Boile and P ssure Vessel Code and appli able Addenda as rcqu red by 10 CPR 5 , Sec ion 50.55a(g), cept wh re specific wri ten relief s been granted HRC rsuant to 10 C 50;ection 50.55a((6)(i).Additional i cctions hall be performed n certa circumfercnt al pipe lds to provide add tional pr tection against pi whip, w ich could dama c auxili ry and control sy terna.BFK Unit 1 3.6/4.6-13 AMENDMQP'go, p p 6 PAGE 5 OF'

3.6.G 3.6.G.(Cont'd)b.Wl the struct al integrity o any ASME C e Class 2 3 quivalent omponent no conformi to the ab e requir ents, rest e the stru ural integ ty of the aff cted compo t to vithin i limit or solate the ffected co onent from all OPERAB systems.4.6.G.For Unit 1 an a ented inservice surveil ce program shall be peiformed to monitor pot tial'orrosive effect of chloride res ue released uring the March 22, 75 fire.The augmented ervice surveillance ogram ia specified as f lovs: a.Brows Ferry Me cal Ma tenance Instruct 53, date eptember 22, 19 5, paragra 4, defines the liquid p trant examinatio required during the f st, second, third and four refueling outages follovi the fire restoration. b.ma Ferry Mechanical Mai ce Instruction 46, dated ly 18, 1975, hppendix defines the liquid pens t examinations re uired during the sixth refueling outage folloving the fire.restoration. BFH Unit 1 3.6/4.6-14 AMENDIHBIT lS 80 6 PAGg JOL 0 s 8%3.6.H.ggg~During a modes of operatio , all snub crs shall be OPE except s noted in 3.6.8.1.All sa ty-related snubber arc 1 ted in Plant Surve llance Instructio l.ith onc or morc snubber(s) inope ble on a system that i required to be OPERABLE n the current plant ondition, vithin 72 ho s replace or restore e inoperable snubber(s) o OPERABLE status perform an engineer evaluation on the a tached compon t 02 decl rc thc attache system inoperable folio the appropri e Limit ng Condition statement for that system.4.6.8.~S Ea safety-related snubber s 11 be demonstrated 0 RABLE by performance f thc folloving a gmcnted inservice inspect on program and thc rcquir nts of Specificatioa .6.H/4.6.8. These snubber are listed in Plant Survci ance Instructions c 0 hs scd in this s cification,"typ of ubbcr" shall me snubbers of the c design and manu acturer, irrespective capacity.2~V Snubber are categorized as ina essiblc or acces ible during re ctor ope tion.Each o these ca egories (inacc ssible accessible) y be nspectcd inde cndently according to c schedule dctermincd Table 4.6.H-1.e visual inspecti interval for each t c of snubber shall be de crmined based on the riteria provid in Ta e 4.6.8-1 and e first i paction intcrv 1 termincd usi this criteria shal bc based upon the pr ious inspection nterval as establis d by the requir cnts in effect before amendment No.210 BFH Unit 1 3.6/4.6-15 NEMMENT IN, 2 go PAG~~ Ol 4.6 H mShhma 3~sual inspec ons shall verify that 1)the snubber has no vis le indications of damag or impaired OPBRhBI TT, (2)atta ents to th fo tion or su portiag st ture are ctional, (3)fast rs for the tachment o the snubber o the comp ent and to the snubber an orate are functio.Snubb rs which appear operabl as a result f visua inspe tions s 1 be cia ified cceptable and be reels ified acceptable or the purpose of establi ing the next visual i paction interval provided that (1)th cause of the e)ection is clearly estab shed and r edied for t particu r snubber for other s bbers espective o type that may be generi ally susceptible; and (2)the affected sn ber is functional y tested in the as-found ondition and determi d OPBRABLS per Specif ation 4.6.5.h revie and evalu tion shall be p rformed documented to)ustify co inued operation wi an unacceptab snubber.If continued peration c ot be gusti ied, the snu er shall declared inope ble and the MITIHG COHDI IOHS POR OP TIOH shall be met.BFH~Jnit 1 3.6/4.6-16 AMB1BEEtlTHg, 2 IP PAGE

C75 Z.b,h q,g, H 4.6,8~SggZZa 4.6.8.3 (Cont'd)hd itional , s bbers a tached sc ions of afety-re ated systems that ve cxp rien ed un ected potenti lly amagi transi ts ines t last insp tion eriod hall be eva ated for thc poa ibil y of c cealcd d e d func onally tested if appl cable, o confi OPBlhB ITf.Snub rs vhi have b cn mad inopcra e as re lt of expected tr ients isolate d e, o other r om events, hen thc rovisio of 4.6..7 and 4 6.8.8 e been m t and other appropriate co rectivc action implem ted, sh 11 not be counte in determining e next isual inspection interval.BFH Unit 1 3o6/4.6-17 .AMENOMENt'lN. 2 To 4.s ldll m ing ch refueling outage a represen tive sampl of 10K of e total of e ch type of saf ty-related ubbers in us in the pl t shall be f ctionally ested either in place or n a bench test The repre tative s le selected or functi 1 testing shall incl e the variou configura iona, opera ing enviro ents, and the ange of si e and ca city of s bbers vithln the types.e representat e sample should be ighed t include m e snubb rs from severe s ice ar as such as near cavy e ipment.The roke se ing and the secu ity of steners or attachment the sn hers to the corn onent an to the snubber chorage all be verifie on snubb rs select for FUH IOKLL TBSTS.BPH Unit 1 3.6/4.6-18 AMENDMENT NO.2 10 pAG a~OF~ Cl, 0, lAN 1g 1ggg 4.6.H.~S u i~<<i 5.C 0 C ter The s bber CT OHAL TEST hall" erif that: a.Activa ion restraining aetio)is chieved in b th t ion and corn ressi vithin the sp cified range, capt t t ine tia dep dent, celer ion lim ing echani al snubb rs may be tes ed to ve fy only at activ tion takes place in oth dire tions of ravel.b.Sn ber bleed or re ease vher required, i present i both c mpression and t ion ithin the pecifi d ange.c.For mech ical sn bbers, the for requir to initiat.or main ain motion of the s bber is not g at eno to overs ress the attached pipi or comp nent dur therma movement, or indica e impendi fa ure of e snubber.d.r"snubbe s specific lly equired ot to disp ace under co tinuous lo the abi, ity of the nubber to vit tand load ithout displa ement shal be, verified.BFH Unit 1 3.6/4.6-19

~6.8.4.6.8.5 (Con d)e.cating met ds may be used to mc ure parameters indircctl or parameters other th those specified if thos ,results can be correl ted to t e speci cd par eters thro estab ishcd me ds.6.kn cngin ring eva tion shall b made of ch failure to mee the FUR OKhL TEST accep ce crit ia to dete e thc ause of the fai ure.The result of this ysis s 1 be used, if plicable in select nubbers be teste in thc subscqu lot in effort to detcrai c the OPE ILITY of other ubbers v ch may bc sub)t to the e failure mod.Sclecti of snubbers fo future te ting may a o bc b scd on th failure lysis.or each s bber that does ot meet t e FUKCTIO TEST ace ptance criteri , an addi onal lot equal o 10 pere t of the rcaa cr of t t type of snu ers shall e functions ly te tcd.Tes ng shall ntinue un 1 no additi 1 noperable snubbers arc found vithin s sequent lots or all snubber of the orig 1 HJKCTI TEST typ have been teste or all susp ct snubbcrs iden fied by the failure ana sis have bc tested, as applicable. BFH Unit 1 3.6/4.6<<20 AMENDMgPNy ~>>PAGE~OF~

4.6.H.u b 4.6.H.6 (Co'd)any snubb selected or functio 1 testing either fai to lo up or fails move,.e., frozen i pl'ace, he cause vill be valuat and if caused y manu cturer or desig defici cy, all snub ers of e same des gn subj t to the s e defec shall be ctiona y tested is tes ing requi ement shall b independ nt of the r uirements stated abov for snubb rs not mee ng the CTIONAL TE accept e criteria.he discov of loose'r missi attachmen fastener vill be e luated to dete ine vheth r the cause ay be loc ized or gener c.The r ult of the valuation ill be use to sele other su pect snu ers for v rifying e attachme t stener , as applica le.7.S a be S Fo the snubber s)found i operable, an engineerin valuation sh ll be perf med on the compo ents which are restrained y the snub er(s).The purpo of this engineer g evaluati n shall to dete ine if the component, restra ed by the BFN Unit 1 3.6/4.6-21 AMENOMBfTNg f63

VAJA>ivs8~4.4.6.8.7 (Cont'd)8~snubber s)vere a ersely affect by the operability of th snubber(, and in orde to e sure that he restrained corn onent rem's capable of , m ting the signed service.u o a 0 S S b Snubbers hich fail he visual inspect on or the FUNCTI AL TEST ac eptance'rite a shall be repaired or r laced.Re lacement snu ers and sn bers which ha e repairs v ch might a fact the FUN TIONAL TEST esults shall meet the FUNCTIONAL ST criter before inst llation i the unit.The e snubbers shall have met e accepta ce criteria subsequent to their most re ent servic , and the FUNCTI NAL TEST m st have been erformed v hin 12 mont s before b ng installed in he unit.9.Permanent or other mptions from vis al inspecti ns and/or unctional t sting for i ividual sn bers may be gr nted by th Commission if justifiabl basis for ex ption is p sented and if applicable nubber life destructive sting vas perform to qualify snubber OP BILITY for the appli able design conditio at either the BFN Unit 1 3.6/4.6-22 AMENDMQP gP.y 8 g Cl 0, iJAN i 9 1989 4.6.H.4.6.8.9 ont'd)completi of thei fabrica ion or at subsequent date.Snubbers exempted shel continue o be liste in e plant i structions ith fo notes in cating the tent of t exemption 10.S e o a C The se ice life o snubbers may b extended b ed on an eva ation of th records of~FU TZONAL TEST m intenance hi tory, and nvironmenta conditions to which the s bbers have been expos d.BFN Unit 1 3.6/4.6-23 elemMM 50.18 8 0 0 JAN i9 1888 THIS PAGE INTENTIONALLY LEET BLANK BFH Unit 1 3.6/4.6-234 AMENOMENT NO.1,6 8 PAGE~OF~I W

~~~~~~~I~~~~W~~~~I~~~I~~~~~~t~I~~~~I I I~Il~g~~~.~~~'I~~~~~~~~~~~~~~~'~~~~I~I'~~~~~~I~~~~~~~~~I~~~~'~~~~~~~~,~~~~~~~~~~~~~~~'~~~~~~I AI~~~~~~~~~~~~~I~~~~~~~~~~~~~A~~~~~~~~~~~~~I~~~~: 'I~~I I I~'~~~I~~I~~~I~~~~~~~~~~~~~~~~'I~~~~~~I~~~~~~~I~~'~~I~~I,~~~~I'~~~*~~~~

Table 4.6.8-1 (Continued) JOt.0sm SHUBBER VISUAL IHSPECTIOK IHTERVAL Hote 4: If e number of una eptable sn hers is e 1 to or ess the nuaber in lpga B but rester the num r in Co A, the next pection i terval 1 be the arne as previous inte al.ote 5: f the number of unicceptab snubbers s equal t or, great than the number in Column , the next pecti interval shall be tvo-rds of th previous erval.ovever, i the number of una ceptable ubbers is 1 s than t e number Column C, bu greater the numb r in Col B, the ext interval s ll be red ced proporti lly by terpolat on, that is, e previo interval ll be red ced by a actor that is e-third the ratio o the diff ence be en the number o unaccep ble snubbers found dur the pr ious interva and the timber in Col B to differ e in the number in Col B and C.te 6: The provisions of Specification 1.0.are applicable for all inspection intervals up to and including 48 months.BFH Unit 1 3.6/4.6-23c AMENOMENT IL 2IO PAGE

t Section 3.4, Reactor Coolant System (RCS)Bases The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content of the proposed Browns Ferry Unit 2 Technical Specification Section 3.4, consistent with the BWR Standard Technical Specification, NUREG 1433.The revised Bases are as shown in the proposed Browns Ferry Unit 2 Technical Specification Bases.BFN-UNITS 1, 2, 5 3 Revision 0 pAGE~oF LI UNIT2 CURRENT TECHNICAL SPECIFICATIPN MARKUP Cl 0 LIMITIHt COHDITIOHS FOR OPZRATIOH P,l URVEILLAHCZ REOUIRZmHTS 3.6 S 0 4.6 S~S 0 cab Applies o the operating status of the rc tor coolant system.~0~v To assure the integ ity and safe operation of the rca or coolant system.Applies to the periodic examination an testing requirements for c reactor coolant system.~Oa~v determine thc condition of th reactor coolant system and thc operation of thc safety devices related to it.t 0 LCo 3.g,g 1.The average rate of reactor coolaat temperature change during normal heatup or cooldom shall not exceed 100 F/hr vhca averaged over a oae-hour period.sm 3.4.9.(1.)During heatups and followed~parameters shall be recorded and reactor coolant~pl temperature determined attic=minute intervals un s ve readings each given lo tion are vithia F.g.4 l a earn Dome Pre sure Convert to pcr vessel rcgi temperatur )Reacto bottom drain tcmpc ature Re rculation 1 ops B d Reactor ve sel bo tom head tern erature e.Rcacto vessel shell adjacent to shell flange BFE Unit 2 3.6/4.6-1 A>na~

2.During all operations with~o a critical core, other than for low-level physics tests, ezcept when the vessel is vented, the.reactor vessel shell and fluid temperatures shall be at or above the temperature of curve 03 of Pigure 3.6>>1.s Reactor vessel metal temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell ad/scent to shell flange shall be dc at least every~<'2 minutes during inscrvice hydrostat'ic or leak test hea the vessel pressure is>3L?psi@.3.Daring heatup by nonnuclear means, except when the vessel is vented Lco or as indicated in 3.6.k.4, during cooldown folloving nuclear shutdown, or during lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve 02 of Figure 3'-1 until removing tension on the head stud bolts as specified in'6.i 5~cst spec ens'eprcseat the reactor vessel, e veld, and veld he t affected zone metal hall be installed in e reactor vc scl a aceat to th vessel all at the re midplane level.Th number and type of ecimens vill be in cordance vith repor KDO-10115. e specimens shall m t the intent of ASIAN E 85-82.<<~Jf Unit 2 3 6/4 6-2 PAGE~AMENDMENT NO.17 0

5 c icAF>0 SEP i 3 1995 ss a 4~gCo g.A SR 3.g.9.I,~oR z.5.l QO 3qq The beltline region of reactor vessel temperatures during inservice hydrostatic) or leak testing shall bc at J or above thc temperatures shown on curve 41 of Figure 3.6-1.The applicability of this curve to these tests is extended to nonnuclear heatup and ambient loss cooldovn associated vith these tests only if the heatup and cooldovn rates do not exceed 15 F per houre The reactor vessel head bolting studs may be partially tensioned (four sequences of the seating pass)provided the studs and flange materials are above 70'F.Before loading the flanges any morc, the vessel flange and head flange must be greater than 82 F, and must remain above 82 F vhile under full tension.4~Si2>06 l No~]Hl p~,pop sg 5.4.9 2 Sg K.iJ.).S d Na4.I S'g?.q.'t-g +Uo~sg s.g.)7 p No+5.When the reactor vessel head bolting studs are tensioned and the reactor is in a cold condition, the reactor vessel shell temperature haaediatcly belov the head flange shall be P o~H C~p e<<dc5 gc.K~g.gq g (pg PAGE OP BFK Unit 2 3'/4.6-3 AMENDMENT NO.2 3 9

gy'le.~~~+It i f~~%t~~~~~~g~~~~'~4~~IAI~','i~eP a 0.TIVE~'%~~~A~'~LT~MWEW.~ XS,At~~~~A~~~~~~,~~o~f~~~~~: I~~~~~~~~~~~~~~~~~~~~~~.~~~~~~~~~~~~~~-'~~~~~~A~~~~A~'~~~~~IA~~I~~,~~~A~~~A~~0 I~~~~~I I~~~~A~~~~~~~~~~.~~~A~~~~~~ 0 ~AI 4.6.E.~Je RmSm~c4 Gw~ifi<4ia~ waar 6"~~gu 4~[gl57$3,'I 2 See Z~S]"Ia.h'~4-Ct~)~8~~Isis Z.q.i 2~Whenever there is recirculation flov vith the reactor in thc SThRTUP or RUH Node and one recirculation pump is operating, thc diffuser to lover plenum differential pressure shall be checked daily and the differential pressure of an individual get pump in a loop shall not vary from the mean of all get pump differential yrcsaures in that loop by more than 10K.3:6.F 4'.F.1.The reactor shall not bc operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if onc recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH vithin 24 hours unless the loop is sooner returned to service.1..Recirculation pump speeds shall bc checked and logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed yump may not be opened unless the speed of thc faster pump is less than 50K of its rated speed 2.Ho additional surveillance required.3 When the reacti5mis not-4n the RUH mode, REhCTOR POWER OPERATIOH vith both recircu-lation pumps out-of-service for to 12 h'ing such interval, restart of 5g, the recirculation pumps ia g.], t.'f pcrmittcds provided the loop gq<c>aischirge temyerature is vithin 75 F of the saturation temperature of the reactor sg 3.0.R.'t 3.Before starting either recirculation pump during REACTOR POWER OPERATIOH/Al c s rgc tcmperaturc and dome saturation temperature. BPH Unit 2 3.6/4+6-12 AMENOMBlT g6.2 2 g PAGE~.OF Q s~)Sic.$(u~3.1 7 NR i 8 1993 3.6.F L~3.g,Q,f, gott 2 vessel water as determined by dome pressure.e a e apse t e in natural circulation and one pump operation must be no greater than 24 hours.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.J 5tc d~gk4j~~'c~ 40~c~r~Isi5 gc/, I 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of'ny ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a COLD SHUTDOWN CONDITION or less than 50'F above the minimum temperature required by NDT consider-ations, until each indication of a defect has been inves-tigated and evaluated. Inservice inspection of ASME Code Class 1, Class 2, and Class 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, Section 50.55a(g except where specific written relief has been granted by NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.'dditional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.II i'FN Unit 2 Stc.WHsgllcr,$ ~@g Cw5 gg g/qg Cg 3 6(4 6 13 AMENOMBF gy, p 06 PAGE~OF 0 ~rrarrrr okra%Sar~aaemIIaaaaaa ~aNIRIIaaarar aaaiislHrararr ~aaHIllirr rrrsllflrraraa aasSISaaaaaa ~rt)rlISrrraaa ~rQRRsaaaaaa ~allasaraaaaa ~aIIRSaaaaaaa E%Ãisrraaarr ~Ilasaaaaaaa 5I5!3~Ksa~~~..~s I I~~.~~~I~~~fl '4 W 3.6.B.1.PRIOR TO ST and at steami rates less than 00,000 Ib/hr, c folloving limits shall apply.4'.BE 1.Reactor coolant shall be c tinuously monitored or conductivity. except vhen there is no fuel in the reactor vessel.a.onductivity, pmho/cm at 25 C 2.0 Chloride, ppm 0 a.Whenever'he continuous conductivity monitor is inopcrablc, a sample o reactor coolant 11 be analyz for cond tivity every 4 rs except as 1 tcd belov.If the, cactor is in COLD SHUTDOWN COHDITIOEg a'ample of reactor coolant shall be analyzed for conductivity every 8 hours.b.Once a veek t e continuous nitor shall bc ected vith an in-li flov cell.This i ine condu ivity cal ration ahall be p formed every 24 ura vhenevei the reactor coolant conductivity is>1.0 pmho/cm at 25 C.2.At steaming rates greater than 100,000'b/hr, the folloving limits shall apply.a.Conductivity, idaho/cm at 25'C 1 0 During startup prior to pressurizing thc reactor above atmospheric pressurcg measurements of reactor vatcr polity shall be performed to ahov conformance vith 3.6.5.1 of limiting conditions. b.Chloride, ppm 0.2 BFS Unit 2 3.6/4.6-5 AMENpMENT RI.224 PAGF/OF~ P I'4~re rs'~war r I'I~v rr~~~-r<Tl? (~VF ttttl~a 1IhtrJttulLeLC)K~Llt%+I'0'k'6 I I~',.l'J:0~~~~~~~~I I~~~~~~I~I~I~~~~~~~~~'~I~~~~~~~II~~~~~~~'~~~~~~I~II I~~~~~~~~~'~~~~~~~~~~I~I I~~~~~~~~~~~~~~~~~~~~~~~I~~11 I~I I~~rr I~~I~II~~~~~~l'~I~~~~~~~~I~~l M~~~I~~I~~~~~~~~~~I I I~'I~~~~~~~~~I o

3.6.B.Cao an ist 4.6.B.Coa em st 4.When th reactor is not pr ssurized vith fuel in the r actor vessel, exce t dur the STARTUP CO TIOH, th reactor vater sha be intained vithin t olloving limits.a.Conductivit 10 pmho/at 25'C 4q enever the re ctor is not pressurized v h fuel i the reactor essel, a sample of e reactor coolant all be ana zed at lea every 96 ho rs for c cdcctivity, loride ion ontent and pH.b.Chlori-0.5 ppm c.pH s ll be betve 5.3 and Se6.0 5.When the time lim ts or um conduct ity or oride conc tration imits are ceded, an orderly shu ovn shall be initiated ediately.The reactor 11 bc brought to the COLD SHUTDOWH COHDITIOH as rapidly as.cooldovn rate permits.5.During equilibrium pover operation an isotopic analysis, including quantitative measurements for at least I-131, I-132, I-133, and I-134 shall bc performed monthly on a caolant liquid sample.6.ene e reactor is critical, thc limits an activity concentrations in the reactor coolant shall not exceed the equilibrium value of 3.2 pCi/gm of dose equivalent I-131.S<d'fdic~~<FiCV T<aM p'~A C RAhl g~~St h/lg Q~'~Y'4<s st=<7, g 6.Additional coolant samples shall be taken vhencver the rcactar activity exceeds onc percent of thc equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are.met: BFH Unit 2 3'.6/4'-1 AMENOMENT NL 2 24 pAGE 0 p if 0 <~$: 3.c.E/Mdiv JUN 2 8)PE 3.6.B.4.6.B.o 3.6.B.6 (Cont'd)This limit may bc exceeded folloving pover transients for a maxim'f 48 hours.During thfa activity transient thc iodine concentrations shall not exceed 26 pCi/gm vhcnever the reactor is critical.Thc reactor shall not bc opcratcd morc than SX of ita yearly povcr operation under this exception for thc equilibrium activity limits.If the iodine concentration in thc coolant exceeds 26 pCi/gm, the reactor shall be shut dovn, and the stem line isolation valves shall be closed immcdiatcly. X6C~TIE'ICWnoN wc~<~"~~R'4 814 Is~s 3.'/.6 i~~(g zpcvaQ 4.6.B.6 (Cont'd)a.During the STARTOP COHDITIOH b.Pollovtng a signfficant pover change**c.Folloving'n fncreaac in thc equflfbri~ off-gaa level exceeding 10,000 pCf/sec (at thc steam get air ejector)vithin a 48-hour period.d.Whenever the equilibrium iodine limit apecifi'cd in 3.6.B.6 is czcecded.Thc additional coolant liquid samples shall be taken at 4 hour intervals for 48 hours, or until a stable iodine concentration belov the limiting value (3.2 pCi/gm)is established. Hovcvcr, at least 3 consecutive simples shall be taken in all cases.kn isotopic analysis shall bc performed for each ample, and quantitative measurements made to determine the dose equivalent I-131 concentratio 7.Wh there is o fuel in thc r actor vesa , technic 1 pecificat n reactor oolant chemiatry fmits do ot apply.there is no fuel in thc eactor vcsa , aampli of reactor c ant ch try t technic specific ion frequency ia not rcquir d.**Por the purpose of this sectfon on sampling frequency, a significant povcr cxchu~e is defined as a change exceeding 15K of rated pover in less than 1 hour.BFH Unit 2 3.6/4.6-8 AMENOMEgl'ltd. 2 p g PAGE V OP l~ CTS R.C,g/+gg J NR i 8 Igga 3.6.F 3.6.F.3 (Cont'd)vessel water as determined by dome pressure.The total elapsed time in natural ci.rculation and onc pump operation must be no greater than 24 hours.Sce 34$4iPic4)jo~ peg W RFIV ISTIC 3.'f./Peci~c l 7 LooPS OpC~4~z i~/hit~ec4o~4.The reactoi shall not be operated with both recirculation pumps out-of-service while the reactor is ia the RUN mode.Followiag a trip of both recirculation pumps while in the RUN mode, imnediately initiate a manual reactor scram 3.6~G 4.6.G The structural 'egrity of ASME Code Clas 1, 2, and 3 equivalent components shall be maintai d in accordance with Spec'cation 4.6.G through t the life of the plant.a.ith the structural integrity of any ASME Code Class 1 equivale component, which is art of the primary sys not conforming to the above requirem s, restore the structural integrity of the affected omponent to within i.ts imit or maintain the react coolant system in either a COLD SHUTDOWN CONDIT N or less than 50'F above the minimum temperature req red by NDT consider-at'ons, until each indicatio a defect has been inves tigated and evaluated. Inservice inspection ASME Code Class 1, Class , and Class 3 components 11 be performed in acco ance with Section XI of t ASME Boi.ler and Pressure V esel Code and applicable A eada ae required by 10 CFR 5 , Section 50.55a(g except wh e specific written relief s been granted by NRC pure to 10 CFR 50, Section 50.55 (g)(6)(i). dditional inspections shall be performed on certain ci.rcumferent 1 pipe welds to prov e addi.tioaal prote ion against pipe w p, which could d gc auxili,ary control systems.BFN Unit 2 3.6/4.6-13 AMENDMBfT%7.2 0 6 pAGs+o~~

c~a.c.g/VC g.MAR I 8 1993 3.6.G 3.6.G.l ont'd)With the struc al integrity of any ASME de Class 2 or 3 equivalent omponent not conformi to the above requir ents, restore the struc al integrity of t'ff ted component to w in it limit or isolate t e a fected component fr m all 0 ERABLE systems.BFN Unit 2 3.6/O.6-l4 AMENOMENT NO.2 0 6 PAGE

During all mo es of operation, all snubber shall be OPERABLE except as oted in 3.6.8.1.All saf y-related snubbers are 1 ted in Plant Surv llance Instructions. 1 With one or more snubber(s) inoperabl on a system that is re ired to be OPERABLE in e current plant co ition, vithin 72 hours replace or restore th inoperable snubber(s) t OPERABIS status and erfora an engineer evaluation on the a tached component or decl re the attached system inoperable and follov the appropriate Limiting Condition statement for that system.Each safety-related snubber shall be demonstrated OHHtkBLE by performance of the folloving, augmented inservice insped'tion program and the requirements of.Specification 3.6.8/4.6.8. These snub rs are listed in Plant Su eillance Instructions. I ks used in this specification,"type of snubber" shall mean snubbers of the same design and aanufac er, irrespective of ca city.2~V Snubbers are ategorixed as inaccess le or acceseibl during reactor operatio.Each of these categor cs (inaccessible and ac essible)may be inspe ted independently according to the schedule determined by Table 4.6.8-1.The vi inspection interv for each type of ber shall be determined ed upon the criteria rovided in Table 4.6.8 and the first inspection nterval determine using this criteria shall be based upon th previous inspection interval as established by the requirements in effect'efore amendment No.225 BFH Unit 2 3.6/4.6-15 AMENOMEHT R0, p 25 PAGE~O~~ 4.6.H.3~0 BPS Unit 2 3.6/4.6-16 Vis inspections shall v fy-that (1)the abber no visible indi tions of dsaae or iepai ed OPERLBILITY; (2)attachments to c foundation o supporting structure e functional, and (3)teners for the etta'cha t of the snubber to th component and to the snub r anchorage are f tional.Snubbers vhich a ar inoperable as a esult of visual inspectioni shall b classified unacce able and aalu be reclassif ed acceptable for e purpose.of establis the next visual insp ction interva.provided t (1)the cause of the r/ection is clearly establ hed and reaedied for t particular snubber and or other snubbers irr pective of type t be generically susceptible; and)the affected snubbe is functionally sted in the as-found co ition and detezained PERABLE per Specific ion 4.6.8.5.k review evaluation shall be per ormed and documented to)tify continued ope tion with an unacceptable snubber If continued operatio cannot be/ustified, th snubber shall be decla d inoperable the LIMITIEG COSDITIOHS R OPBIRTIOH shall be me OMENS NJ;p 25 pAGE~OF~ 0 4.6.8.3 (Cont'd)~k tionally, snubbers tached to scctio of safety-related s tees that have experienc unexpected potentially aaaglag transient since the last inspect n period shall bc evalua ed for the poss ility of concealed d e and functionally tc tcd, if applicable, to nfirm OPBRABILITY nubbcrs which ha been made inoperable thc result of un ected transients, solated daaage, o other random events, en thc provisions of 4.6.7 and 4.6.H.have been ct and any o r app opriate corre ive ac ion implemen d, shall not be counte in determining c next visual inspection terval.BFH Unit 2 3.6/4.6<<17 NENOMENT Ng.p g 5 PAGEOS~

4.6.H.~S~~s~4~ing each refuel tagc, a reprcs ative sample of 10K o the total of each type o safety-relet d snubbers in use in thc lant shall be functio ly tested either in pla or in a bench test.The representative sample a ected for functional eating shall include the various configuratio operating environm ts, and the range of size and capacf ty of scc srs sithin the types.rcprescntat e sample should be eighed to.include orc snubbers from severe crvicc areas such as n r heavy equipment. stroke setting d thc security of fast rs for attachment of th snubbcrs to thc compon and to the snubber anch age shall be verified o snubbcrs selected or FUKCTIOKlL TESTS.BFH Unit 2 3.6/4.6-18 AMENDMENT go.p p 6 Gp~dOP~~ 0 JAN 18 1888.6.H.5.e snubber FUHCTIOHhL TEST shall verify that: 4 a.hctivation (re raining action)is a ieved in both te ion and compress n vithin the specif ed range, except tha nertia dependent, a eleration limiting echanical snubbers may be tested to verify only that activation takes place in both directions of trav b.Snubber bleed, r release vher required, is present n both compress n and tension vithin he specified rang o c.or mechanical snubbers, the force required to initiate or maintain motion of the snubber not great enough to overstress the a ached piping or comp ent during, the movement, or to ind ate impending failure f the snubber.d.Fo snubbers specifically quired no t to displace under continuous load, the ability of the snubber~to vithstand load vithout displacement shall be verified.BFN Unit 2 3.6/4.6-19 AMENDMENT NL l 6 9 PAG~~r't:~8' C C7S S.g.g 6.g P JUL 055%.6 H.4.6.8.5 (Cont'.eating methods may be used to measure parameters indirectly or parameters other than those specified if those results c be correlated to the specified param ers through estab shed methods.6.ka ineering evaluation 1 be made of each failure meet the PORCTIOEhL TEST acceptance criteria to determine the cause of e failure.The result this analysis shall be, if applicable, in sel cting snubbers to be t ted in the subsequent lot n an effoxt to determine th OPERABILITY of other snub rs which may be sub)ect the same failure mode.election of snubbers for f ure testing may also be bas on the failure sis.For each snubber t does not meet the CTIOKAL TEST acceptance criteria, an additional lot equal to 10 percent of remainder of that type f snubbers ahall be fun ionally tested.Testing s 1 continue until no dditional inoperable snubb s are found vithin subsequ t lots or all snubbers of e original tOICZZOELL tyya hare been)tested or all suspect snubbers identif d by the failure analys s have been tested, as applicable. BFH Unit 2 3.6/4.6-20 ~AIH<n~an e.p pq PAGE~iOF/ 4.6.H.4.6.H.6 (C'd)If any snubber selected for functional test either fails to croup or fails to e, i.e., frozen in ace, the cause vf.ll b valuated and if caus by manufacturer or d ign deficiency, all snubbers of the same design subject to the same defect shall be functionally tested.This testing requirem shall be independen f the requirement tated above for s ers not meeting e FUNCTIONAL TEST cceptance cri,teria. The discovery of loose or missing attachment fasteners vill be evaluated to determine vhether the cause may be.localize or generic.The res of the evaluation'l be used to sel other suspect ubbers for veri ng the attachment teners, as applicable. 7.For the snubber(s) und inoperable, an e neering evaluation s be performed on the co nents vhich are restra ed by the snubber(s). Th urpose of this ineering evaluation shall be to determine if the components restrained by the BFN Unit 2 3.6/4.6-21 NmoMen R.I SO waco~~ JAN f 9 1888 4.6.H.4.6.H.7 t')snubber(s) e adversely affecte y the inoperability of snubber(s), and in orde ensure that the restrained component remains capable of meeting the designed servic 8~u b s Snubbers v fail the visual inspect or the FUN NAL TEST acceptance c teria shall be repaired or replaced.Replacement snubbers and snubbers Which have repairs vhich might affect the FUNCTIONAL T results shall meet t FUNCTIONAL TEST teria before instal ion in th>>unit.Th snubbers shall have m the acceptance cr ria subsequent to their st recent service, and the FUNCTIONAL TEST must have been performed vithin 12 months before being install in the unit.9.V ua anent or other exemptions from visual inspections and/or functional testing for individual snubbers y be granted by the C ssion if a justifiable asis for exemption is esented and if appli e snubber life dest ive:esting v performed to qualify snubber OPERABILITY for the applicable design conditions at either the BFH Unit 2 3.6/4.6-22 @t)MENT NO.X 6 0 PAGp~o~

4..H.4.6.8.9 (Cont'pletion of their fabrication or at a s sequent date.Snubbers so empted shall continue be listed in the plant tructions with footnotes icating the extent o the exemptions. e service life of sn bees may be extended base on an evaluation of the ecords of FUNCTIONAL TEST maintenance h tory, and environment conditions to vhich the ubbers have been exp sed.BFN Unit 2 3.6/4.6-23 AMENOh)BIT ND.X 6 0 PAGp~50F~

'JAN 19$888 THIS PAGE IHTEHTIORALLY LEFT BLANK BEE Unit 2 3'/4.6-23a NENtjMENT NO.X 6 0 r s a~~koF~S ble 4.6 H-1 SHUBBER SUAL IHSPECTIOH IHTERVAL Population or Catego Notes Column A Extend Interval Column B Repeat Int al 4 Column C Reduce Interval e Hote 1: The next visual inspection interval or a snubber population or category size shall be determined ased upon the previous inspection interval and the n er of unacceptable snubbers found during that interval.ubbers may be categorized, based upon their accessibility d ing pover operation, as accessible or inaccessible. These tegories may be examined separately or)ointly.Hovever, e licensee must make and document t decision before any pection and shall use that decisio as the basis upon wh to determine the next inspection erval for that categ Hote 2: Interpola on between population or category siz and the number unacceptable anubbers is permissible Use next lover integ for the value of the limit for Col A, B, or C if tha integer includes a fractional value unacceptable snu ers as determined by interpolatio Hote 3: If the number of unacceptable snub rs is equal to or less than ,the number in Column A, the nex inspection interval may be twice the previous interval not greater than 48 months.BFH Unit 2 3.6/4.6-23b FINED'8~(

T le 4.6.8-1 (Continued) S BBR VISUAL IHSPBCTIOH IHTERVAL Hote 4: If th umber of unacceptable snub rs is equal to or less the number in Column B but eater than the number in Co A, the next inspection terval shall be the same as e previous interval.Hote 5 If the number of unacce able snubbers is equal to or greater than the number in Co C, the next inspection int al shall be two-third of the previous interval.How er,'if the number of unacce able snubbers is less than the umber in Column C, but eater than the nuaber in Col , the next interval be reduced proportionally by terpolation, that is, previous interval shall be r ced by a factor that is e-third of the ratio of the d ference betveen e number unacceptable snubbers found uring the previo interv 1 and the number in Column B o the difference the numbe in Columns B aud C.Hote 6: The provisions of Specificatio 1.0.LL are applic le for all inspection intervals up to including 48 mon BPH Unit 2 3.6/4.6-23c AMENOMENT RQ.p Zg p.<gE~$'F~3' UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF g

'n LIMITIHt COHDITIOHS FOR OPERATIOH SURVEILLAHCE REQUIREMEHTS

4.6 Applies

t the operating s atus of the res or coolant syst plies to the p iodic ination and t ting req rements for th reactor cool t system.To assure the integr and safe operation of the react r coolant systeme To dete the condition of the reactor oolant system d the operation f the safety devices relate to it.1.The average rate of DO S't.g reactor coolant temperature change during normal heatup t or cooldoma shall not exceed 100 F/hr shen averaged over a one-hour period.ill!<Mgcooldoms, the c ggAQ,)following parameters shall be recorded and reactor coolant temperature determined 36 ute intervals til 3 succ s ve readings a each given lo tion vithin F.a.team D e Press e (Conve to upp vess region t erature)b.actor ttom drain empera re Reci culatio loops A B d.eactor v ssel bottom head tern erature e.React vessel shell ad)a ent to shell fl e BFS Unit 3 3.6/4.6-1 PAGE~pp~ 3.Daring heatap by nonnuclear means, exccyt when the vessel is vented or as indicated in 3.6.h.4, daring cooldown folloving nuclear shutdown, or daring lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve$2 of Figure 3.6-1 antil removing tension on the head stad bolts as specified in 3.6.L.5.LCo 3A g 2.During all operations vith a critical core, other than for lov-level pbysics tests, LCg.except vhcn thc vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temycratare of carve 03 of Figure 3.6-1.5 3'.M..eactor vcsscl metal temperature at the outside surface of the bottom head in thc vicinity of thc control rod drive housing and reactor vessel shell ad)accnt to shell flange shall be recorded at least every LZ minutes daring ervice hydrostatic or test~cn the vesee 5g g,q,fj,)~ressurc is>312 psig 3.Test specimens reyresent the reactor vessel, b e veld, and veld hea affected zone metal I be ins alled in th reactor v sel aQ ent to the cssel at the c c mi plane level.The be and type of sp im vill be in acc vith GE reyort lKDO-10 15.The specimens shall meet the intent of hSTN E 185-82.BFH" Unit 3 3'/4'-2 PAGE OP~AMENDMENT NO.14 1 l.i <Pc ca'P i 3 1995 5$'3A.g,)NnH'.g, 4.The beltline region of reactor vessel temperatures LC.O (during inservice hydrostatic 9'g.9 or leak testing ahall be at or above the temperatures shown on curve¹1 of Figure 3.6-1.The applicability of this curve CD Chese.Ccats is cztended to ncmnuclear heatup and aabient loss cooldom associated with these tests only if the heatup and cooldem rates do not exceed 15 F per hour>>SRg.g.g.l Nog)p)y p~g sC.9.'L9.2.~R~~9~9e (>>+bloke, sg z.q,9.a+.Mom LC.b 3 9.1 5.The reactor vessel head bolting studs may be partially tensioned (four sequences of the seating pass)provided Che sCuds and flange.IaCerials are above 70 F.Before loading the flanges any more, the vessel flange and head flange must be greater Chan 70 F, and Itust reaain above 70 P whfle under fuX1 tensicm.5.When the reactor vessel head bolting studs are tensioned snd the reactor is in a Cold Condition, the reactor vessel t eratur imediately below'he head flange shall be Prod~w~c~Pr 5',g,),5>PAGE OF BFK Unit 3 3.6/4'-3 AMENDMENT Ng.y 9 8

ecifim'3 I~+@i 6.The pump in an idle recirculation loop shall not be started unless thc temperatures of thc coolant vi in thc idle o erat ng recirculation loo are w 5D F of each other.+3'.o9.6.PJo)r urfag ST of an idle rec rculation loop, the temperature of the reactor coolant the operat and die loops l bc+y o c I Co$,4,cj 7.The reactor recirculation pumps shall not be started unless the coolant temperatures betveen the dome and bottom head drain are vithin 145 F.Prior to atartiag a recircu ation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall bc compare M3 p~~>~Pcy IeN$p, g~~PAGE OP~BPS Unit 3 3.6/4.6W

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4.6.E.~Je~~I Sc<'Sus+t Ei'rc4onQ~ c~ey+~>F'N iSTS 3q.~I 5<@5gs+iCi ca/ion Q~cha<g~<kr'P'hl l5T5 SI I 2.Whenever there is recirculation flov wit the reactor in the STARTUP or RUN Mode and one recirculation pump is operating, the diffuser to lover plenum differential pressure shall be, chcckcd daily and the differential pressure of an individual jet pump in a loop shall not vary from thc mean of all jet pump differential pressures in that loop by more than 10'.3.6.F at 0 4.6.F e 0 tio 1.The reactor shall not bc operated vith onc recirculation loop out of service for more than 24 hours.Pith the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITXOH vithin 24 hours unless thc loop is sooner returned to service.1.Recirculation pump speeds shall be checked and logged at least once per day.sR9.g 2.Folloving onc-pump operation, the discharge valve of the lov speed pump may not be opened unless thc spccd of the faster pump is less than 50K of its rated speed.3.When thc reactor is not in thc RUH mode, REhCTOR POWER OPERhTIOH with both recirculation pumps out-of-service for up to 12 hours i ermitted Dur ng such interval restart of thc recirculation pumps is permitted, provided the loop discharge temperature is vithin 75 F of the saturation temperature 2.Ho additional surveillance required.~R 3.Before starting either recirculation pump during REhCTOR POWER OPE&Bd'e loop dis arge temperature and dome saturation temperature. pAGE~OF~BFH Unit 3 3.6/4.6-12 AMENDMgfT NP.y 8$ .6.F 3,9A,9~Ivo&2.of the reactor vessel water as determined by dome pressure.The a e apse t e n natural circulation and one pump operation must be no greater than 24 hours.See V<S+Pi~gon kr C$r Bpe Isis r,v.j 4.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor's in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50 F above the mintunun temperature required by HDT consider-ations, until each indication of a defect has been investigated and evaluated. 1.Znservice inspection of ASME Code Class 1, Class 2, and~Class 3 components shall be performed in accordance with Section XX of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.3ietifiaaiion 6r Cjgunyg for C T5 y,i, p./q BFH Unit 3 3.6/4.6-13 AMENOIHEgf gg, y p g A8~$OF~~ .8 V.S.Z JUN 2 8 199$3.6.B.1.PRIOR TO ARTUP and at steam rates less t 100,000 lb/hr the following limi s shall apply.4.6.BE l.eactor coolant shall be continuously monitored for conductivity except when there is no fuel in the reactor vessel.a Conductivity, pmho/cm at 25 C 2 b.Chloride, ppm 0.1 I~I a.Whenever ghe continuous onductivity monitor inoperable, a samp of reactor cool shall be ana zed for c ductivity every hours except as listed below.If the reactor is in COLD SHUTDOWH COHDIT105, a sample of reactor coolant shall b analyzed for conductivi every 8 hours.b.e a week the ontinuous monitor shall be checked with an in-line flow cell.This in-line conductivity calibration shal be performed eve 24 hours whenev the reactor coo t conductiv is>1.0 ymho/cm t 25iC, 2.kt steaming rate greater than 10 ,000 lb/hr, the fo owing limits shal apply.a.Conductivity, pmho/cm at 25'C 1.0 2.During st tup prior to pressur ing the reactor'bove tmospheric pre ure measurements o reactor water quality hall be performed to show conformance with 3.6.B.1 of limiting conditions. b.Chloride, ppm 0.2 BPK Unit 3 3.6/4.6-5 AMENDS@~0.I 8~AGE/OF Ci CTS Z.(,8 q.a.g OEC 0 7 1994 4.6.B.3~ht steaming rates greater than 100, 0 lb/hr, the react water quality exceed Specifi tion 3.6.B.2 only'r the time limits pecified belov.Exc eding these time limi or the folloving maximum ality limits shall be caus for placing the re tor in the COLD OWH CORD IOH.Conductivity time above 1 idaho/at 25iC-2 ve/year.Limit'0 o/cm at 25 C 3~Whenever the reactor is operating (including HOT SThHDBY COHDITIOH) measurements f reactor vater quali shall be performed ccording to the foll ing schedule: a.oride ion content d pH shall be measured at least once every 96 hours.b.Chloride ion content sha be measured a least every 8 h rs vhenever reactor conduc vity is>1.0 o/cm at'C.b.oride concentration ti e above 0.2 p 2 vecks/ye Maximum Lim Oo5 ppme c.sample of reactor coolant shall be measured for pH at least once every 8 hours vhenever the reactor coolant conductivity is>1.0 idaho/cm at 25iC.c.The rea or ahall be placed in the SHUTDOWNS COHDI IOH if pH<5.6 or>8.6 for a 24-hour period.BPH Unit 3 3.6/4.6-6 N>NEON.I 86 PAGE

3.6.B..6.B.Coo t st 4.When the eactor is not pre urized vith f 1 in the re tor vessel, cept duri the SThRTUP HDITIOH, the eactor vater 11 be mai tained vithin the fo loving limit.Conductiv ty-10 pmho cm at 25 4.Whenever t e react r is not pressuri d vith uel in the rea or ves 1, a sampl of the eactor cool t shall e analyzed at east ev 96 hours f conduc vity, chloride on conten and pH.b.Chio de-0.5 c.p shall be tveen.3 and 8.6 5.Wh the tim limits or co ctivity or oride c ncentration limits ar exceeded, an ordirly hutdovn shall be initia ed immediately. The react r shall be brought to the OLD SHUTDOWH COHDITIOH as apidly as cooldom rate permits.5.During equilibrium pover operation an isotopic analysis, including quantitative measurements for at least I-131, I-132, I-133, and I-134 shall be performed monthly on a coolant liquid sample.6.enever e reactor is critical, the limits on activity concentrations in the reactor coolant shall not exceed the equilibri~ value of 3.2 pCi/gm of dose equivalent I-131.S<c 3'u.~f;~/on QP.C~ge<t S~nt isis s,g,g Section 6.hdditional coolant samples shall be taken vhenever the reactor activity.exceeds one percent of the equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are met: BFH Unit 3 3.6/4.6>>7 AtaENoMENT go.y 8 y PAGE

3.6.B.3.6.B.6 (Cont'd)This liait Nay be exceeded folloving, pover transients for a aaxiarm of 48 hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gw vhenever the reactor is critical.The reactor shall not be operated aore than 5Z of its yearly pover operation undir this exception for the equilibrium activity liaits.If the iodine concentration in the coolant exceeds 26 pCi/ga, the reactor shall be ahut dovn, and the steaa line isolation valves shall be closed iwaediately. See V~q6<i'm~n $r.~~~n>s Vs r.q,(.;~;, Scc ho g 4.6.Bi 4.6.B.6 (Cont'd)a.During the STARTUP CORDITIOK b.Folloving a significant pover chLnge**c.Follovtng.an increase in the squilibritm off-gas Xerel exceeding 10,000 pCi/sec (at the steaa)et air e)ector)vithin a 48-hour period.d.Whenever the equilibrhm iodine liait specified in 3.6.B.6 is exceeded.The additional coolant liquid saayles shall be talon at 4 hour'ntervals for 48 hours, or until a stable iodine concentration belov the liaiting value (3.2 yCi/ga)is established. Hovever, at least 3 consecutii saaples shall be taken in all cases.An isotopic analysis ahall be perfozaed for each sawple, snd quantitatire aeasureaents aade to deteraine the dose equiralent I-131 concentration. 7~the e is fuel in the actor esse , te cal pecif cati reac r coolant cheai try 1 ts do not apply.7.en ther is fuel e reac r v sl)saapl of r ctor c olant cheai ry a tschni 1 spec ficat on fra ency is t requi ed.For the purpose of this section on saapling frequency, a significant pover exchange is defined as a change exceeding 15Z of rated pover in less than 1 hour.BFN Unit 3 3.6/4.6-8 NENBMgpgQ, z gz'AGE~OF~~ tl .6.F 0 a 3',F,3 (Cont'd)of the reactor vessel water as determined by dome pressure.The total elapsed time in natural circulation and one pump operation must be no greater than 24 hours.see swiJ"cqAo* k, C+~~B~H>~T~3'~~Pgcirculq~ 9 p s a+His 5ecti on 4.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.~6.G S U 4.6.G The struc ral integrity of ASME Code Cla 1, 2, and 3 quivalent compon s shall be intained in acc dance with S ecification 4.6.G hroughout t life of th plan~a.With the s ctural int rity of any Code Class equival t component, which is pa of the pr system, not onforming to e above re irements, res ore the s ctural inte ity of the ffected compo ent to with its limit or intain the reactor coo t system i either a Cold Shu own conditi or less t 50'F abov the min temperatu e~requir by HDT cons der-atio , until each ndication of a defect has be n investigated and evaluated. 1.nservice pection f ASME Code Clas 1, Class , and Class 3 omponents hall be perform d in accor ance with Sectio XI of the ASME Boiler and P essure Ves el Code d appl cable Add a as re ired by CFR 50, ction 50 55a(g)ex ept where ecific itten r ief has b grant by HRC rsuant to 0 CFR 50, Section 0.55 (g)(6 (i).2.Additio inspect ons shall be perform on cert in circum rential ipe welds to pr ide addi ional prot tion aga nst pipe whip, whi could d ge auxiliary and control s stems.BFH Unit 3 3.6/4.6-13 AMENDafEHT NO.r 7'9 PAGE~OF~~ a vs 3.6.g./q, 0.4-NY31m 3.6.G 3.6.G.l (ont'd)4~6.G b With t structu al integ ity of ASME Co Class 2 or 3 equ alent co onent no co forming t the abov r quirement , restor the tructura integrity of the affected omponent o vith its lim or isol e the affect d compon from l OPE syst BFH Unit 3 3+6/4+6-l~ AMENDMENT NQ.g 7 g PAGE~OF/ 0 3.6.8.4QQhhCXk 4 6 8 2m8hma During all modes operation, all saubbcrs sha be OPERABLE except as noted n 3.6.H.1.hll safety-rcl ed snubbers are listed in lant Surveillance astructions. 1.With o or more snubb (s)inoperable a a sys em that is re red to b OPERhSLE in cur cnt plant cond ion, wi in 72 hours r place o restore the i operablc ubber(s)to 0 RABLE~tatus and per orm aa engineeriag c aluation on the atta cd component or declare e attached system ino erable and f olla appropriate Limiting Condition statement for that system.Each sa ety-rclatcd snubber shall e demonstrated OPE by performance of e folloving augmented i rvice inspection program thc requirement of pecification 3.6.4.6.8.These snubbers ar listed ia Plant Surveill e Instructions ka use in this speci icatioa,"type of snub er" shall mean sn bcrs of the same d ign and manufac rcr, respective of c acityo 2~Snubbcrs are ategorizcd as inacccss le or accessible uring reactor operation Each of these categori s (inaccessible and ac ssible)may inspe ed indepcnd tly acco ing to the s cdulc det rained by Tab e 4..8-1.The v ual pection int al for each type of ubber shall be determin based upon the criteri provided in Table 4:6.-1 aad the firs iaspecti interval determi ed using this criter a shall be based upon c previous insp ction interval as est lishcd by the req ircmcnts in cffcct before amendmcat No.183 BFH Unit 3 3.6/4.6-15 IIILENOMENT No, y 88 PAGE~OF

Visual inspectio shall verify that (1)e snubber has no visibl indications of Cease o layaired OPERABILI~(2)attachR s to the founda on or supporting struc re are functional, and 3)fasteners for the at chment of the snubber t the component and to the ubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified unaccep able and aay be reclassif d acceptable for e purpose of establish the next visual insp ticm interva.provided t (1)the cause of the r ection is clearly establi ed and reaedied for t particular snubber and or other snubbers ir spective of type that be generically susceptible; and (2)the affected snubber is functionally tested the as-found conditicm detezained 0 per Specification 4.8.5.reviev and ev uation shall be perforae snd docuaented to gustif continued operatio with an unacce able'snubber. If ccmt ued operation cannot be Justified, the snubber shall be declared inoperable and the LINZTIEC COHDITIOKS FOR OPERLTIOK shall be aet.BP5 Unit 3 3.6/4.6-16 AMENDMENr ND.Z 8g PAGE 9 pp/g 4 6 8 Bmhhern 4.6.8.3 (Cont'd)kd tionall , snubb rs a tached sectio of afety-r ated s teaa that have erience unexpected poten illy d iag tr eats s ce the ast ins ection riod 11 be ev lusted or the ssibili y of c cealed e func onally tested if app cable, to conf i OPB LITT.Snab ers vhi have be sad inoper le as th re t of expected t ansien , isolat aaage, r other oa events when the rovisions of 4..8.7 and.6.8.8 ve be met and other ap opriate rrectiv a ion 1ap1 ented, 1 ot be co ted in deterain the n t visual inspect interv 1.a BFH Unit 3 3'/4.6>>17 AMENOMBfT NO.I 8 3 PAGE~oF~S

4.6.8.4.Daring ch refue ng outage a repres tative aaapl of lOX the total of e ch type aa ty-relat snubbera in e in the ant shall be unctional teated either in place r in a ben teat.The re eaentativ aLRple aele ed for fun ional tes ng shall 1 elude the va ious confi rations, o crating en ironILenta, and e range o size and capacity snubbers ithin'he type.The repres tative s pie should e veigh d to inclu e aore ubbers frc seve e aervi e areas such aa ear he equfpaent. The atro e setting the aecuri of fastener for atta ent of the bbera to e coaiponent to the anu ber anchorag shall be ve fied on anu era selected for CTIOKAL TESTS'FR Qnit 3 3.6/4~6-18 em0mHrHU X83 PAGc~oF-4.6.H.S u e 5.snubber CTIOHhL ST shall v rify that: a.Activ ion (restr ning=acti)is achie d in oth tension and co pression vi in the ecified r e, except hat inerti dependent, accelerati limiting mechanic snubbers may be test to verify only t at activation takes place in bot dir tions of trav l.b.S ubber bleed, r elease where equired, is present i both compression and tension within the specified range+c.For me ical snu ers, the rce require to ini ate or mai ain mo ion of the ubber is t great eno to verstress e attached piping or omponent during t rmal movem t, or to i icate imp ing failur of the snu er.d.For nubbers sp ifically re ired not t displace er continu load, he ability f the snubbe to vithst load vithout displacem t shall be verified BFH Unit 3 3.6/4.6-19 NENDMENT NO.1 3 4 e Ga~l'r~8' 0' 4.6.8.4.6.H.S (Co?Lt'd e Testing methods may be used to mLcasurc p raactcrs indirectly or aactera other than sc spccif ic if those suits can be correl ed to the aye iad parameters ough established aethoda.6 X~a&a BFS Unit 3 tel 3 6/4.6-20 ka engineeri evaluation shall be de of each failure to ecc e PUECTIOKAL TEST acce ance criteria to d I%inc the cause of the ailure.The result of this analysis shall be used if applicable, in aele snubbera to be t ted in the subsequent lo an effort to deterainc OPERhBILIIT of other an era which say be sub/a to the aaae failure Selection of snubbera r future testing aalu also be aaed on the failure analysis.For each ubber that does not aee the lUKCTIOML YES acceytance criteria, ddltional lot equal to yercent of the reaa r of that type of snub ra shall bc functi ly tc ed Tcsti?Lg s ontinue until no itional inoperable anu era are found within suba cnt lots or all snubbcrs the original tUKCTZ TEST type have bean i teat or all suspect ambbers identified by the failure analysis have been teated, aa apylic able.NENOMENT N.y 83 pAG<~Ocean l8'

19 1989~6 oHo'gt~4.6.H.6 (C d)If any snubber ected for functi testing either ils to lockup or ls to move, i.e., ozen in place, the caus will be evaluated and if caused by manufac er or design defici y, all snubbers the same desig bject to the s defect shall be ctionally tested.This testing requirement shall be independ t of the requireme stated above for bbers not meetin e FUNCTIONAL TES cceptance criteria.e discovery of loose or missing attachment fasteners will be aluate to determine w er the cause may b ocalized or generic e result of the luation will be u to select other suspect snubbers verifying the achment fasteners, applicable. 7~ed 1 b nts r the snubber(s ound inoperable, gineering evaluation all be perform on the c ponents which are restr ned by the snubber(s The urpose of this e ineering evaluat shal o determine if e compone restrained b he BFN Unit 3 3.6/4.6-21 N ENDMQPNP Zeg PAGE~~~P~ cTS 3.4.'JAN 19 1"-P.6.H.~~~s 4.6.8.7 (Co d)snubber(s) vere ersely affected by e inoperability of the s ber(s), and in order to re that the restrained co onent remains capable of ecting the designed ice.8.S ubb s Snubbe vhich fail the visual insp tion or the CTIONAL TEST accepta e riteria shall be re red or replaced.Rep cement snubbers and s hers vhich have repair hich might affect t FUNCTIONAL TEST result shall meet the FUN OHAL TEST criteria b ore installation i he it.These snubb s shall have met the ac ptance'riteria sub quent to their most rec service, and the FUNCTIO TEST must have been rformed vithin 12 mo before being ins led the unit.C 0~J BPN Unit 3 3.6/4.6-22 Pe ent or other exemptions fr visual inspections d/or functional testing for individual snubbers y be.granted by the C ission if a Justifiable asis for exemption is esented and if applic e snubber life destruc ve testing vas rformed to qualify s ber OPERABILITY for e applicable design conditions at either the NENDMENr NK X ac pAgE OF CVg Z.6 JAR i 9 1988 4.6.H.~t~4.6.H.9 (C'd)completion of thei fabrication or a subsequent date.Snu rs so exempted shall c inue to be listed in t plant instructions with f tnotes indicating th extent of the exemp ns.10.S e ora The se ce life of snubbers may e extended based on aluation of the rec s of FUNCTIONAL TESTS, maintenance his ry, and environment conditions to vhich the nubbers have been osed.BFN Unit 3 3.6/4.6-23 AMENOMENT NQ.X3 pwaa~~o>~~: Cl 0 mrs.c.a/v<.e JAN I 9 1989 THIS PAGE IHTEHTIOHALLX LEFT BLAHK BFH Unit 3 3.6/4.6-23a ANENOMENT Ng.X'3PAGE~6'F~~ Table 4.6.8-1 SHUBBER VISUAL IHSPECTIOH IHTERVAL Populati or Cate ry 0 Col A Extend terval Column B Repeat Interval Coluan C Reduce Interval Hote 1: The next visual ection interval fo a snubber population category size sha be deterained bas upon the previous inspection inte al and the amber unacceptable snubbers found during t interval.Snubb s aalu be categorized, ased upon their ac essibility during er operation, as aces ible or inaccess le.These categori s aay be exaained sepa tely or/ointly Heaver, the lic ee auat aake and docua t that decision efore auy inspecti and shall use that de sion as the baai upon which to dete e the next inspecti interval for category.Hote 2: Inte lation between po ation or category siz s and the n er of unacceptable ubbera is peraissible. Use next'lower in cger for the value f the liait for Col A, B, or C if t integer includes a fractional value of cceptable ubbers as detera d by interpolation. Hote 3: If the nuaber of cceptable snubbers is equal to or lese than the nuaber in Co A, the next inspect on interval cay be tvice the previ interval but not gre ter than 48 aonths.BPH Unit 3 3 6/4.6-23b AMENDMENT N, g 83 Table 4.6.8-1 (Continued) SSUBBER VISU ISSPECTI05 I hL If e nuaber of cceptable snub rs is equal o or less the number in Colum B but g eater than nenber i Co umn k, the nex inspection in erval shall b the saa as e previous int rval.f the nuaber f unacceptabl anubbera ia e to o.greater than the n~r in Column C the next inap ction in rval shall be tv thirds of the revioua inte al.Ho@er, if the nuaber of cceytable bere ia leaa the er in Column C, t greater the member Golda, the next interval 1 be reduc proporti by int rpolation, that is, the previous terval ahall e reduc by a facto that is one-third of e ratio of e differ ce between e nuaber of unaccepta e anubbera fo dur the previo inte al and the n er in Column to the ifference the n~rs in Col B and C.yrovfakoma f Specificati 1 O,LI re applicab for all inspection int rvala up to including 48 aontha.O'I 3 6I4.6-23c NENOMEgr gp.y 88

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEMPERATURE (P/T)LIMITS MINIST TIVE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore,'nderstandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. ~A~Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.These surveillances are a duplication of the regulations found in 10 CFR 50 Appendix H.These regulations require licensee compliance and can not be revised by the licensee.Therefore, these details of the regulations within the Technical Specifications are repetitious and unnecessary. Furthermore, approved exemptions to the regulations, and exceptions presented within the regulations themselves, are also details which are adequately presented without repeating the details within the Technical Specifications. Therefore, retaining the requirement to meet'he requirements of 10 CFR 50 Appendix H, as modified by approved exemptions, and eliminating the Technical Specification details that are also found in Appendix H, is considered a presentation preference which is administrative in nature.A3 For clarity, the terms"prior to and during startup" and"prior to" have been replaced with"15 minutes".This Frequency is effectively the same since the proposed Surveillance now must be performed no more than 15 minutes prior to startup of the idle recirculation loop.This is essentially equivalent to the current requirements. I BFN-UNITS 1, 2, 8E 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEHPERATURE (P/T)LIHITS A4 e Proposed SR 3.4.9.4 requires verification that the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature are within 50 F of each other.CTS 3.6.A.6/4.6.A.6 requires verification that the temperatures between the idle and operating recirculation loops are within 50 F of each'ther. The temperature of the"operating recirculation loop" is considered equivalent to the RPV temperature. Therefore, this change is considered administrative. Proposed SRs 3.4.9.5, 6 5, 7 require the reactor vessel flange and head flange temperatures be verified>82 F, while CTS 4.6.A.5 requires the reactor vessel shell temperature immediately below the head flange be recorded.The BFN procedure that implements this requirement requires the vessel flange and head flange temperature be verified and requires the shell temperature be recorded.Since the intent of the surveillance is to verify vessel flange and head flange temperature to satisfy CTS 3.6.A.5 and both the current and the proposed SRs do this, the two are considered equivalent. As such, the proposed change is administrative. TECHNICAL CHANGE-NORE RESTRICTIVE A new Surveillance Requirement has been added.SR 3.4.9.2 ensures the RCS pressure and temperature are within the criticality limits once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality. This is an additional restriction on plant operation. H2 Three new Surveillance Requirements have been added.SR 3.4.9.5 ensures the vessel head is not tensioned at too low a temperature every 30 minutes.SRs 3.4.9.6 and 3.4.9.7 ensure the vessel and head flange temperatures do not exceed the minimum allowed temperature. These are additional restrictions on plant operation since the current requirements have no times specified. M3 ACTIONS have been added (proposed ACTIONS A, B, and C)to provide direction when the LCO is not met.Currently, no real ACTIONS are provided.These ACTIONS are consistent with the BWR Standard Technical Specification, NUREG 1433, and are additional restrictions on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE~OF~

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEMPERATURE (P/T)LIMITS TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Details of the methods for performing Surveillances, and any, requirement to record data, are relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs."Specific" The Frequency of this Surveillance has been changed from 15 minutes to 30 minutes.Verification that RCS temperature is within limits every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes is reasonable in view of the control room indication available to monitor RCS status.Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. In addition, this new Frequency is consistent with the BWR Standard Technical S ecification, NUREG 1433.Ll P'I L2 The Frequency of this Surveillance has been changed from 15 minutes to 30 minutes.The metal temperature is not expected to change rapidly due to its large mass, thus a 30 minute Frequency is adequate.In addition, this new Frequency is consistent with the BWR Standard Technical Specification, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 0 t-JUSTIFICATION FOR CHANGES CTS 3.6.B/4.6.B -COOLANT CHEMISTRY'ELOCATED SPECIFICATIONS Rl The chemistry limits are provided to prevent long term component degradation and provide long term maintenance of acceptable structural conditions of the system.The associated surveillances are not required to ensure immediate operability of the reactor coolant system.Therefore, the requirements specified in current Specification 3.6.B/4.6.B did not satisfy the NRC Final Policy Statement technical specification screening criteria as documented in the Application of Selection Criteria to the Browns Ferry Unit 2 Technical Specifications and have been relocated to plant documents controlled in accordance with lOCFR50.59. Revision 0 PAGE JUSTIFICATION FOR CHANGES CTS 3.6.G/4.6.G -STRUCTURAL INTEGRITY RELOCATED SPECIFICATIONS Rl The structural integrity inspections are provided to prevent long term component degradation and provide long term maintenance of acceptable structural conditions of the system.The associated inspections are not required to ensure immediate operability'f the system.Therefore, the requirements specified in current Specification 3.6.G/4.6.G did not satisfy the NRC Final Policy Statement technical specification screening criteria as documented in the Application of Selection Criteria to the BFN Unit 2 Technical Specifications and have been relocated to plant documents controlled in accordance with lOCFR50.59. BFN-UNITS I, 2, 8L 3 Revision 0 peCiE~OF~'

JUSTIFICATION FOR CHANGES CTS 3.6.H/4.6.H -SNUBBERS TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Snubber inspection requirements are part of the BFN Inservice Inspection (ISI)Program and are being relocated to the ISI program documents. Requirements for the ISI Program are specified in 10 CFR 50.55a to be performed in accordance with ASIDE Section XI.NRC regulations contain the necessary programmatic requirements for ISI without repeating them in the proposed BFN ISTS.Changes to the ISI Program are controlled in accordance with 10 CFR 50.59.With the removal of operability requirements from the Technical Specifications, snubber operability requirements will be determined in accordance with Technical Specification system operability requirements. BFN-UNITS 1, 2, 5 3 Revision 0 Section 3.4, Reactor Coolant System (RCS)Bases The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content of the proposed Browns Ferry Unit 2 Technical Specification Section 3.4, consistent with the BWR Standard Technical Specification, NUREG 1433.The revised Bases are as shown in the proposed Browns Ferry Unit 2 Technical Specification Bases.BFN-UNITS 1, 2, Ea 3 PAgp evi sQF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP C 1 4 0 LI TIHC COHDITIOHS FOR OPERATIOH SpE'Cgi QLQo.,)NOV 22l988 SURVEILLAHCE REQUIREMEHTS 3.5 4,5 0 G Applies o the operational status of the core and containmen cooling systems.Appli to the surv llancc reqair cats of the rc and containm t cooling s terna when the corre pending limi ng condi-tion for o cration is i effect.To assare the OP ILITY of the core and conta ament cooling systems under all c nditions for which thf.s cooling c pability is an essential respons to plant abnormalities To verify the PERABILITT of the core and contai cnt cooling systems under al conditions for which this cool capability is an essential response to plant abnormalities. L,CD 1.The CSS shall be OPERABLE: 3iS.l (1)PRIOR TO SI'JLBTUP from a COLD COHDITIOK> or (2)when there is irradiate fuel in the vessel and when the reactor vessel prcssure is greater than atmospheric prcssure, except as specified in Specification 3.5.h.2 Ll SR ax.l"i a-rage CC~'4+SR3eS I J C IlI Simulated Automatic Actuation test~i~u~c Once/I8~, Oyera&ag SR3~6b.Pump OPERA-BILITY Per Specifi-cation 1.0.MM Co tor Per Sycc+i-Op rate atioh, l.~Val e OPE ILI 1.Core Spray System Testing.5g P,S,],6 d.System flow Once/W rate: Each loop shall dclivcr at least 6250 gpm against a system head corres-ponding to a BFH Unit 1 3.5/4.5-1~AMENDMENT NO.15 9 FA+', sF t5 I S fcc,'0'c~k'on 3.S.l AUG 02 t989 5'4 B.S.i.4 105 psi differential prcssure bctveen the~reactor vessel and the primary containment. 2~+Cog)a)Aires csc e OPERAB rs gL/Sc>'n/Node 3 in Iahr If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days rov/t5 a ac ve components in the other CSS loop and thc RHR stem c hal V ve er ccif cati n MM S'g P f l.2..Once/Verify that each valve (manual, povcr-operated, or automatic) in the infection flovpath that is not locked, scaled, or other-vise secured in position, is in its correc g7 positione 3.If Specification 3.5.A.l r Specification 3.5.A.2 c ot RCIToN bc met, the reactor shal be 84 H'laced in the COLD SHUTDOWH COHDITIOH vithin ho s.S6 ca 4.Shen the reactor vesse pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERABLE pump and associated diesel generator shall bc OPERABLE, except vith the reactor vessel head remove s ccific as spccificd in 3.5.A.1.2~Eo Idigio s+gl~e r ited Sc'c X~l'*echo~ ger Agre'>>SPH tsTS S.t.l Except that an automatic valve capable of automatic return to its ECCS position vhen an ECCS signal is prcscnt may be in a position for another mode of operation. BFH Unit 1 SCe 34SHCfca4o~ P,~fjgggS BvH isis y,s.~3'/4,5-2 PAGE~OF~hMENDMENTNO. 16 9

The RHRS shall be OPERABLE fP.(1)PRIOR TO STARTUP from a COLD CONDITION; or 2)when there is irradiated fuel in the reactor vessel and when the reactor vessel prcssure is greater than atmospheric, except as specified i'Specifications 3.5.B.2 ou h 3.5.B I.a.SRZ.S.t S b.C~GRAS.I 4 d-L.l AI4l oi Simula ted Automa tie Actuation Test Pump OPERA-BILITY Motor Opera-ted'valve OP ERAB ILITY Pump Flow Rate O ce/,s 3 Per Specification 1.0.Per Specification 1.0.MM Once/9 mes4hc 5g35 i g fO Test Check Per Valve Specification l.O.MM c/e Verify that cc each valve (manual, power-s operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, ig i correct position.Zld~w'R 3.5.l.g/4oK SR3,.S,I.V g.Verify LPCI ce/subsystem cross-tie valve is closed gull power removed f rom valve operator.BFN Unit 1 Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5W Ex ept hat an a orna c val e apab of a to-mati retu to its ECC posit on w en an CCS s gnal is pr sent y b in a posit n f r an her ode of o e tio AMENOMENT gg, 2Pg PAGE ~k'cia'~Hen P,S;/AUG 02 tGGG 2~Vith thc reactor vcsscl pressur less 105 psig, the may be emoved from s rvicc (cept hat tvo RHR p ps-cont en cooling mode d asso atcd eat cx ers t r ia OPE)fo a pe iod not to exceed hour vhile b dra ed of s press n cham er qu ity vatcr fille vith primary coolaa quali y vater rovide that ring coold vn tvo oops th o e pump er lo or o loo vith tvo p ps, an associate diesel generatorsy ia e core ray stem are OPE s RS.XI Each LPCI pump shall deliver 9000 gpm against, an indicated system pressure of 125, psig.Tvo LPCI pumps in thc same loop shall deliver 12000 gpm agaiast an indicated system pressure of 250 psig.2.An air test oa the dryvcl aad torus headers and nozzles shall be conducted once/5 years.A vater test may bc performed on thc torus hcadcr in lieu of the air test.Se'e 3usRFicogon P<QQQC'5 gi BFN~5'4'ig 3.If e RHR um (LPCI mode)is inoperable, the reactor~may remain in operation for a period not to exceed 7 days AcT>NJ H r cma RHR pumps (LPCI mode)and both access paths of the RHRS LPCI made)and the CS PERABLE.5 CC Su5AQ<a~~At Ch~g&~BFu tSTS Z,S,l Ll 4.If 2 RHR um s (LPCI mode)become inoperablc, the reactor shall bc placed in the CO SHUTDOWH COHDITIOH vi thin hours.3'c.Qei Jn Noh'ia l2hrs Rnol BFH Unit 1 3.5/4.5-5-a.-L~ne~g I NENDMENTNO. 16 9 0 Qpccl gicahon 3,5;)8.ACT loQ5 B 4H.If Specifications 3.5.B.1 thro h are t an e reac or shall be placed in the COLD SH WN CONDITION within hours.Pb LZ, When e pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. inMQs 3"~5~q Sec'&Cicahora~M~S 4w 8p J4 l5TS 7>>$.%9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.10.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling re not required 10.No additional surveillance required.BPN Unit 1 When there xs irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capabi.lity is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5>>7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.NEHOMENT NO.204 PAG~

1989 12.If one RHR pump or associated. heat exchanger located the unit cross collILectio the ad)a t unit is i operable fo any reason (i eluding val e inoperability, pip break, etc , the reactor may emain in op ation for a eriod not exceed 30 day provided remaining RHR pum and associ ed diesel generato are OPE 13.If RHR cro s-connection lov or heat remov capability lost, the unit may remain in ope tion for a period t to exceed 10 days unless su capability is btoredo 12.Ho additional surveillanc required.13.Ho additional surveillance re uired.SC B.S..14.1 reci culati n pump di charge valves shall be P PRIOR TO S (or close if permi ted el ether in the speci cations)14.All recirculation pump discharge valves shall be tested for OPERABILI. during any period of gg p g (,S COLD SHUTDOWH COHDITIOK exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.BFK Unit 1 3.5/4.5-8 AMENDMBlr 80.y 69 PG 7 oF Is SF'c;'on 3.5.I 5 c r~s+,'<'mt'on Q~C4+eS@r l~TS.St.cHon 3.3.g.(St.4 Sus+4 ta fjon 5a~c's foe BFn)lSTs 3.V.7 4.9.4.4.(Cont'd)Ce d.'4-kV a own board voltages shall be recorded once every 12 hours.The loss of voltage and degraded voltage relays vhich start thc diesel gcncratora from the 4-kV shutdown boards shall be calibrated annually for trip and reset and the measarcmcnts logged.These relays shall be calibrated as specified in Table 4.9.k.4.c. 5.Logic Systems a.Coamon accident signal logic system is OPZRhBLE.>NSti Acegy~P g Qe BC'N)Sy>Z,~,l 5.480-V RMV Boards ID and 1E SR 3,5.l.l?LCt~i%ao Once c 0$e automat c transfer feature for 480-V RNOV boards ID and 321 shall be fanctionally tested to verify auto-transfer capability. b, 480-V load ahedd ,logic system is OPERJUKE.6.There shall be a minimum of 35,280 gallons of diesel fuel in each of the 7-day diesel-gcncrator fuel tank assemblies. 5~<~+if':nk'en fv Chr~X+A)f5'Q j f g BPS Unit 1 3 9/4.9-7 NENOMEHT Ng.y 8 g PAGE oF I~ $Pcc,i4'.5'.1 NOV 0 4 199t 3.9.h.d.The 480-V shutdown boards lh and 1B are energized. See Yustihcation far Changes 5r at=~1sT',g,7 e.The units 1 and 2 diesel auxiliary boards are ized f.Loss of voltage and degraded voltage relays OPERhBLE on 4-kV shutdown boards h, B, C, and D.g.Shutdown buses 1 and 2 energized. ~Ce Su544icaHon Par f/~&so~isis z.z.Li 5CC ScaS+kcce'gase Qg Q/IongeS'~ BFN isaac p,g,~h.Th-react r mo or-op rated valve RMO b ards&1E are ener ized, th m tor-ge era r ()ets I, 1D>and lEh se ice 4.The three 250-V unit batteries, the four shutdown board batteries, a battery charger for each batte an assoc ated battery board are OPERhBLE.See g~gqg;cab'on f r Qaga W BPH 1575 3.f.g a~d'tt.7 4.Undervoltage Relays a.(Deleted)h.Once every 18 conchs, the conditions under which the loss of voltage and degraded voltage relays are required shall be simulated with an undervoltage on each shutdown board to demonstrate that the associated diesel generator wi'll start.PAGE BFH Unit 1 3.9/4.9-6 AMEHOMEHT tbtO I 8 6 NOV 18 1888~I 12.When one 480-V ahutdovn board is found to be ZEOPERhBLE, the reactor vill be placed in HOT STANDBY COHDITIOS vithin 12 hours and COLD Q93TDO OIITI01 vithin 24 hours.13.If e 480-V ard ag se is ZEO REk R 0 05 tinue for a eriod t o exce sev days proteid the eaa 480-V ard sets and ir so cia lo 0 4.Zi aag 480-RMV aS sets ecoa Z50 rea tor shall pla in the COLD S CO TIO vithin 24 urs.5.If the reqaireaents for operating, in the coaditioas specified by 3.9.B.1 through 3.9.B.14 cannot be aet, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWNS COHDITIOK vithin 24 hours.See X~5+pjc'Phon 5c Chants fee l5 TS B.f 7 See 345 hPicggon for (4~gy t))-"H)S)S Sac@an).E BFN Unit 1 3.9/4.9-14 PAGE Cn pP l5 AMENDMENT NO.y5 8 SPCC 4'eA 3,5, FEB 0 7 I99I 3.5.D 0 s 4~5.D 1.The equipment area co er asso ated vith each pump the equipment area'coo r associated vith each t of core spray pumps and C or B and D)mu be OPERABLE at all mes en the pump or p s se ed by that speci c coole is considered t be OPE LE.l.Each e ipme t area coole~is opera ed i con)unction vith the quip nt served y that pa ticu r cooler;erefore, e e uipment a a cooler are ested at the same freq ency s the pump vhich th se e.2.When an equ ment area cooler, is not PERABLE, the pump(s)se ed by that c er must be co idered inop ble for Tec ical Specific on purposes.c.co 1.3.s.(Wr'f('CR4l ~7 (ll]P(ep5eck~ok 4~$R3,5,(,g The HPCI sy em s all be enever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWH COHDITIOH or as specified in Specification 3.5.E.2.OPERABILITY shall be deter-" mined vithin 12 hours after reactor steam pressur (oA reaches 150 psig from a COLD COHDITIO r a ernat ve y RI TO ST P usi an 1 ry ste su ly.M3 S.(<r a.copse(hloR br sR3~(,'l c~HPCI Subsystem testing.shall be.performed as follovs:+fu or Simulated Once/18 Automatic months Actuation Test 3 Pump OPERA-BILITY Per Specification 1.0.Moto Oper-Per ted alve Spe fica$io 0 RAB ITY.0.sR 3,g.l,g r~(Oow SR s.s,(.q Flov Rate at Once/4-a monQm ra or e PI3 oe tng pr s re$2o&lolo(s,'g lg BFH Unit 1/t(Q 3.5/4.5-13 AMENDMEHT NO.I S 0 Spec'4'<<a< 3',s, (FEB 0 7 1991 Coolant In e 2~~Bog//<~re/Qfioa p LPCI an OPERABLE.S are VCce P~till~~If the HPCI system is inoperable, the react r may remain in operation a d not to exceed M ays, prov ed t e R Llo K/4S+o'p SR 3.5, i.w sR z.s.>.8 Flov Rate at Once/18 psig months Verify that each valve (manual, pover-operated, or automatic) in the injection flow-path that is not locked, sealed, or othervise secured in Once R3 position is in its correc osition.W The HPCI pump shall deliver at least$000 gpm during each flow rate test.ld~s 3~If Specifications 3.5.E.l~~or 3.5.E.2 are not met, t e reactor vessel pressure~ac s shall be reduced to 150'"'sig or less vithin R4 hours.3&*E cept that an aut matic va ve c able aut mati retu to s ECC posit on wh an ECCS ignal is pre ent may b in a ositio for anothe mode o operation. F.'eacto Co e splat o Coo in F.Reactor Core Iso at o Cooli~~~BFN Unit 1 1.The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall e~X~A;Rc ho<4c'~~5.5/4.5-14 O'C 80%lSYS SaSo3 1.RCIC Subsystem testing shall be performed as follows: a.Simulated Auto-Once/18 matic Actuation months Test NIENOMEMt Hp.X 80 pAQF/~QF~~~

SgeCig'ro,>on g, g.NY 1 9 l994 Qqhcqb's)iQ Six valves of the Automatic Depressurization System shall be OPERABLE: (1)PRIOR TO STARTUP from a COLD CONDITION, or, L ISb (2)whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 485 ps g, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.PAfaScd 4~4 Sg g,S,ll f he relief gal s is ove d 4.6..2)9$'.Durin ea cretin c he following tests shall be performed on the ADS: L(Ac~i Of SR P,g J lg a.A simulated automatic actuation test shall cut~l8~55.be performed PRIOR TO STARTUP te each outa 2~P'GTlodb E WTlog 8+8 3~R CTiDA5 G.With one of the above required ADS valves inoperable, provided the HPCI system the core spray system and the LPCI system re OPERABLE, res ore the inoperable ADS valve to OPERABLE statue within 14 days or be in at least a HOT SHUTDOWN CONDITION within the next 12 hours and reduce reactor steam dome ressure to~495 psig within hours.IS@sc With two or more of the above required ADS valves inoperable, be in at least a HOT SRJTDOWN CONDITION within i2 hours and reduce react steam dome pres'sure to g+95 psig within hours.ISo 3'4, L2, al recpaka ed-.frig s+c(Sg 3,5/3 frofasect AcTloH p Rss~'rM Ps@on L)(LCo 3.s,gg QopB 2 cubi whig 7gry Old DE 3 QPifhirt (3/r PlODE 9 luis n 3'7/lfS LI L I's (W A~<nlrb)BFN Unit 1 3.5/4.5-16 NENOMENT NL 2P g PAGE~3 OF~5 7 S eCigca.'on R,g 0KC O V 894.C 6~C 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours I for the air sampling system..!!The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.See Y~sSka Hen P, Cg~~W BPt4 ls75 g.q,g 3.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.4.e.D 1.When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours.The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. 1.Approximately one-hal of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.+~>4"<wagon 4c CQ~4 SPA)sTs g~,~Pfop6c'd~a4~S.S.i.U LR 2.In accordance with Specification 1.0.MN ach elie va ve shall be manually opened unt ermocoup es and a usti monit s do stre of t e val e ind cate earn i ow fro the ve.BFN Unit 1 3.6/4.e-lo AMENPlHEHT NL 2 Z3~3 PAGE 1 QF I5 Whenever the core spray s stems, LP I, HPCI, or C ired to e OPERAB , th discharg pipi from the pum discharg of the e systems to t e last block v lvc shall be f led.Rig~SC S.S.<,1 e following surveillance requirements shall be adhered to assure that the discharge piping of the core spray systems, LPCI~HPCI, d RCI are filled: Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 8 psig 48 psig S~c~~f'cabin f c Chavez)a f4'~~)STS 3,g.y The suctio of the IC an pumps shall e aligned to the condensate storage tank, and the pre ure supp ession chambe head tank shall no lly be aligned to erve the discharge piping of th RHR and S pumps.Th condensate head t may be used to serve th RHR an CS disc ge piping the P hea tank i unavailable The pressur indicators the discharg of the RHR CS pumps sha indicate not less than liste below.1.ve mont h he RHRS LPCI and Conta ament Spray)and core spray system, the dischar pipin of these s stem Pll v ed om e h'gh oin wa fl d rm ed 2.F ow ng any per od where the LP I or ore s ray s t hav not een r quire to e OP LE, he di charg pi ng of t ino rable syst sha 1 be v ted f m the high in prior o the eturn of the system o service.3.Whenever the HPCI or RCI system is lined up to take suction from the condensate storage tank, the dischar e pipin of thc HFCI an CIC bc v n p t f t s t ob rve on a monthly basis.4.en th RHRS and th CSS are r uired o be OPERAB , the pr sure x dica ors whx h mon or th disc arge li es shall be mon tore daily d the pressure recor ed.PAGE~OF L+BFN Unit 1 3'/4.5-17 AIENOlHENT No.2 05 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 5 ce Sic 4io~'3~~4 0 0 COO G S S S p)NOV 22 1988 LIMITIHG COHDITIOHS FOR OPERATION SURVEILLAHCE REQUIREMENTS 3.5 CO MSHXILS CO COO G 4.5'O CO~SS~~S COO G cab cab t A lies to the operational st us of the core and coat ament cooling systems.Applies to the s eillance requirements of th core and containment cooling ystems vhea the corresponding lim ting condi-tion for operatioa is effect.0 ect ve 0 ect vc To assure the OP ILITY of thc core and conta cnt cooling systems under all co itions for vhich this cooling cap ility is an essential rcsponsc to lant abnormalities. To ve fy the OPERABILITY of the core an containment cooling systems er all conditions for which this ooling capability is an essential zesponse to plant abnormalities. ca o Sec cto t ahe 3.5.l (1)4ppl;((4'.Ig <>>CSS shal1 be OPERABLE: PRIOR TO STARTUP from a COLD CONDITION~ or rhea there is irradiated fuel in the vessel and when thc reactor vessel pressure is greater than atmospheric pressure, except as specified in Specification 3.5.A.2.sR 3.s.l.g a.P,l gQ tle4,*sea.s,l.t act Ao lated Automatic Actuation test g~v~enc cc]e Qpomtekag p"t PRIES I f P3 Pump Opera-Pcr Specifi-bility catioa 1.O.MM c.otor O Op ated pg Valv OPERABILITY Per pec fi-n~en l.h MN 1.Core Spray System Testing.san.s.l.(, d.System flov rate: Each loop shall deliver at least 6250 gpm against a system head corres-ponding to a Once/M~P'~/qz.4~s BFH Unit 2 3.5/4.5-1 AMENDMEHTNO. 15 5

ON g gC,,'fir.qC(o~ ~<I~'S Cont'd)g.s.t.6 105 psi differential pressure bctveen the reactor vessel and the primary containment. P~e.Ch ck Valve Per Speci cati a.o.m 20 PA'ld 9 If one CSS loop is inoperable,.the reactor may remain in operation for a period not to exceed 7 days rovidi all active components in the other CSS loop and thc PCI mode and the diesel generators are OPERABLE.Sc w Ho~IR Qrg Once/each valve (manual, pover-operated, or As automatic) in the~ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc p7 position.3~Ad riOat~+8 4, If Specification 3.5.h.l or Specification 3.5.h.2 cannot be met, the reactor shal e placed in the COLD SHUTDOWH COHDITIOH vithin hours.LQ.When thc reactor vesse pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERJQKZ pump anil associate4 diesel generator shall be OPERABLE, ezcept vith the reactor vessel head remo c ln 3.5.A.or PRIOR TO STARTUP as specified in 3.5.h.l.OA6 addit nal s~eQ lance is r ire 74 s4iflc44s~~Col4<<wg+J g4~a~~(srs~.s.~ccpt that automatic v ve capable f autom tic re rn to its CS posi ion vhen ECCS sign is presen may be in a position or another mode of operation. BFH Unit 2+~~~>4 t,'cA(d~4yi Qd~*-gS/Is~~3.5/4.5-2 PAGE~QF~5 AMENDMENT NO.16 9 oo xng nme t ent)PRIOR TO STARTUP from a COLD CONDITION; or 2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.The RHRS shall be OPERABLE 0.Simulated Automa tic Actuation Test l.a.ce>g Opee~Ig Oyer c.otor pere-er ted valve Specxf xcatxon OPERABIL Sg S.5,).C d.Pump Flow Once/&Rate 6 e.Testa e Check A~Valve Per Specification 1.0.MN S~~<l g b.Pump OPERA-er BILITY'pecification 1.0.MM 5R 35I.'L f Once/04m+h Verify that each valve (manual, po~er-operated, or J/cg automatic) in the injection flow-path that is not locked, sealed, or other~isa secured in pcsi-tion, i in its P'7 correc positron.81 Unit 2 3.5/4.5-4 sR 3.</.2.So@.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. g.Verify LPCI subsystem cross-tie valve is closed nruj power removed from valve operator.nce/AMENPMgfT gg.2 2 P Except t>at an automat c valve capabl of auto-mati return to'ECC position en an CCS signa is present may ve in a position for another mode of operation.

S ci fic~A~w 3.5.I AUG 02 5gg 3 4~nt ainment 2~With the rea tor vessel pressure le than 105 psig, the RHRS'be removed from serv ce (except that tvo RHR pump-containment cooling mode associated heat exch ers must remain OPE LE)for a per d not to ceed 24 hours hile b ng drained of ppression ch er quality vater and fill d vith primary cool t quality vater provi ed that dur cooldown o loops wi one pump per oop or one oop vith tvo pumps, an associated diesel generators, in th core s ra s stem a LZ 0 0~)sR 3.S.ach LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Two LPCI pumps in thc same loop shall deliver 12000 gpm against an indicated system pressure of 250 psig.2.An air test on the dryvell and torus headers and nozzle shall be conducted once/5 years.h vater test may be performed on the torus header in lieu of the air test.c See: a'aA4<J>o<g~~imam r.C.2 3~PcTI0 hl A p,cgw H Lf If one RHR ump (LP mode)is inoperable, the reactor may remain in operation for a period not to cxcecd 7 days rov e e rema n ng pumps (LPCI mode)and both access paths of the LPCI mode an S an t c cse generators remain OPERABLZ Pl~S,ea-s$:AcJ o-4'-Cw~(~~Sir BFN la~<.S.l e l~4.If 2 umps (LPCI mode)become inoperable, the reactor shall be placed in thc COLD SHUTDOWN CONDITION vi thin hours.3b$2.plo6 3 is./2 4rS ance BFN Unit 2 3 5/4.5-5~(.-E~OF lP AMENDMENTNO. l6 9

m Pi 4 tainmen t inmen t 8.Acvlo~l2 6w nce Sce~~s]A'zi4~ 4r C4o~igg ger Bf'N<S~)Rr.~esse pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.0.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required 10.No additional survei'lance required.Unit 2 When there is irradiated fuel in the reactor and the reactor is'ot in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capabili.ty is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.J AMENDMENT HD.2 2 3 5'c i'gic<4io AUG 02 198Sent va S 12.If three RHR pumps or associated heat exchangers located on the unit cross-connection in the ad)acent uni s are operable for any r son (ncluding valve inopc bility, pi e break, etc.), thc r ctor may emain in operation for a eriod not to exceed 30 day provided the remaining RHR pump and associated diesel generator are OPERABLE.3.If RHR cross connection flov or heat removal pability is lost, the unit may re ain in operation for a period not o exceed 10 days unless such c ability is restored 4.11 rec rcu at on ump d charge vcs sh ll bc ERABLE IOR TO STAR (or cl scd if permit d clscv re in these specifi tions).13.No additional sur illance required.gg 3.5.(,5 14.411 recirculation pump discharge valves shall bc tested for OPERABILITY uring any period of sq q~~~COLD SHUTDOMH CONDITION exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.Unit 2 3.5/4.5-8 AMENDMENT NO;1 6 9 WGE TZOH A>S ec',f<e 4'.S.NOV 0~1991 HTS 4 d.The 480-V shutdown boards 2A and 2B arc energized. e.The units 1 and 2 diesel auxiliary boards are energized. See~~xlP+lw cog r g-gem ic.rs 3.8.7 f.Loss of voltage and-degraded voltage relays OPERABLE on 4-kV shutdown boards A, B, C, and D.g.Shutdown buses 1 and 2 energized. The 480 reactor motor-opera d valve (OV)boar s 2D&2E e encr ized wi motor-gen ator g)s s 2DH, 2D , 2EH, d 2EA service.see~~sligicPaw 4l c4~Jag~PpN Ic7s 2.3.8.J S e~us4Ki~k.o. 4r n.-g~4<NFL Isis 3.8.J The three 250-V unit batteries, the four shutdown board batteries, a battery charger for each battery and ssociatc atte board arc OPERABLE.4.Undcrvoltage Relays a.(Deleted)b.Once every 18 months, thc conditions under which the loss of voltage and degraded voltage relays arc required shall be simulated with an undervoltagc on each shutdown board to demonstrate that the associated diesel generator will start.3.9/4.9-6 AMENDMgfTN0,$9 g 4.9.h.4.(Cont'd)QQ, 3~gle ksc41>~phag~Qr LJ~f5+/5CC4 l~~3 8'I cc 4~L4i4<ce4 4t C~~QI-gP'tJ J S T'5 3.8.7 c.The loss of voltage and degraded voltage relays which start the diesel generators from the 4-kV shutdova boards shall be calibrated annually for trip and reset and the meaaurcmenta logged.These relays shall be calibrated aa specified.h.4.c.4-kV shutdown board voltages shall be recorded once every.12 hours.5 Logic Systems a.Comaon accident signal logic system ia OPERABLE.5.480-V RNOV Boards 2D and 2E I 5~is.~P e aut c transfer feature for 480-V RMDV boards 2D and 2E shall bc functionally teated to verify auto-transfer capability. b.480-V load shedding, logic system ia OPERAS.6.There shall be a miniaaun of 35,280 gallons of diesel fuel in each of the 7-day diesel-generator fuel tank assemblies. See r~k l;o<~.Ch.Zc~~gin irrs R,Z.3 BES Unit 2 3.9/4.9-7 NEHDMEHT NO.1.9 I.)s L.Whea onc 480-V shutdovn board is found to be IHOPERABLE, thc reactor vill be placed in the HOT STAIBY COHDITIOH vithin 12 hours and COLD SHOTDOWH COHDITIOH vithin 24 hours.13.If onc 48 V RNOV board mg set is I PERABLE, REACTOR POWER 0 ERATIOH may cont e for a period not to cccd seven daysg pr ided the rema 0-V RMOV board sets and their associa ed loads remain OHGKBLE~I 4.If auy tvo 4 0-V RMOV board mg s s become IHOPE , the rea or shall placed in c COLD S WH CO ITIOH vithin 24 hours.15.If thc requirements for operating in the conditions specified by 3.9.B 1 through 3.9.B.14 cannot be mct, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWH COHDITIOH vithin 24 hours.+C ZML4i C+g~~~h~g 4av QpW I S7 5 5~~Qy~p g BFH Unit 2 3'/4.9-14 AMENDMENT NO.Zg 4 pAGE 0 OF f5

Cl~: livia(g t~Iitl I~~~~~II~~~~~~~I~~~~II~~~~~~~~t I~I~I~~~~II~~~~~~~~~~~~~I~I I I~~~I~~~~~I~III~~~~~\~I t~~II~~~~~~'~~~~~II~~~~~~~I~I~~~~~'I~~~~~~~~~rr'I'MrttE tr 4 I t O'3 I 8 0 4+0!I%~ax r m i I~I II t II~~~r I~~~~~~~~~~~I~~~~I~~~~~~I~~~I I'h~~I~~~~~~~~~~I~~I I~I~'~'~I~~I~.~'~~~~~II~~~>~r ClS' 5'iVi'cc4io~ 3.R/FEB 0 7 t9gt ct on Io c (45 P sip SR S~Flov Rate at Once/18 448-psig months The HPCI pump shall deliver at least 5000 gpm during each flov rate test.SR f~E.C.I.2 Verify that Once/each valve (manual, pover-operated, or automatic) in the injection flov-path that is not locked, sealed, or othervise secured in positio , is in its correc position.Pro Ahviog D 2~If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed I days, rovided the RHRS(LPCI and RCICS are~OP LE vtr nc pcyio~As~3~p,n iota g4H If Specifications 3.5.E.l or 3.5.E.2 arc not mct, t e~o PE reactor vessel pressure shall be reduced to 150 psig or less vithin 3'ours.*ept that au automatic va e capable o autom tic retu to its EC posi on vhen a ECCS signa is present y be in a position or another mode of operation. a o o F.a t C 0 oo BFN Unit 2 The RCICS shall be OPERABLE vhenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 3.5/4.5-14 1.RCIC Subsystem testing shall bc performed as follovs: a.Simulated Auto-Once/18 matic Actuation months Test AMENDMEHT NO.1 9 0 Sec Z~>$;~iicqfjo~ 4~C~ra egal]g7g 5 5'3

FEB 0 7 199t 1.Six valves of the Automatic p,g Depressurisation System shall be OPERABLE: (1)PRIOR TO STARTUP from a COLD COHDITIOH, or, I p)~Q~(2)vhenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than si except in the COLD SHUT-DOWH COHDITIOH or as specified in 3.5.G.2 and 3.5.G.3 belov.(Vg~+(5~rg.Pro~Ny W s~3.s.I.LQ rve~cc o the rel val&s is vercd i 4.6.D.1.Duri each operat c c 0 ov ng tests shall be performed on the ADS 4'R a.A simulated automatic 3 5.(lO'ctuation test shall be performed PRIOR TO STARTUP er eq efuel c 2~P,neo&jEcTlo 4 e~H Pith one of thc above required ADS valves inoperablc, provided the HPCI s stem, e core spray system and the LPCI stem re OPERABLE, restore the inoperablc ADS valve to OPERABLE status vithin 14 days or bc in M least a HOT SHUTDOWH CO5QITIOH vithin thc next lg'nours and reduce reactor team dome pressure to sig vithin hours.ISO e ances P~~~sf 3.s;],g 9 o(-+Ace(ou r=3~Pcyso nE&Vith tvo or more of the above required ADS valves, inoperable, be in at least a HOT SHUTDOWN COHDITIOH vithin 12 hours and reduce reactor team dome pressure to sig vithin hours.58 b L.5 BFH Unit 2 R4'g~~M Ackiy-g.I (C Z u 3.~)Robe 2-1,~'7 ggg vf<I34 s A J7 I rs~g.lZ 3.5/4.5-16 A DEC 0 7 l994 G COND 0 S OR 0.6.C Coo 2~CC I I I4 I age Anytime irradiated fuel fs in the reactor vessel and reactor coolant temperature fs above 212 F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, thc reactor may remain in operation durfng the succeeding 24 hours for the sump system or 72 hours for the afr sampling system.4.6.C Co a a 2.With the afr sampling system inoperable, grab samples shall be obtafned and analyzed at, 1cast once every 24 hours.S~<+'S Ja 4 C J.;O~CO/'M gag fo~SPV'i<7-S~.q.g The afr sampling system may be removed from service for a period of 4 hours for calibration, function testing j and maintenance vithout providing a temporary monitor.3~If the condition in l.or 2 above cannot be met, an orderly shutdovn shall be initiated and the reactor shall be placed fn the COLD SHUTDOWNS COHDITIOS vithfn 24 hours.4s6 D BPÃUnit 2 20 Sec~,t;g;<;P,C<~W QFIIJ IS~P.Q.3 I R'I.S.I.L I OJo4.~Sa 3.s.l.II c8 3.6/4.6<<10 When more than onc relief valve is knovn to bc failed, an orderly shutdovn shall be inftiated and thc reactor 4eprcssurfzed to less than 105 psig vithin 24 hours.The relief valves are not required to be OPERABLZ in the COLD SHUTDOWN Approximately one-half o f all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating cycle.All 13 valves vill have been checked or replaced upon the completion of every second, cycle.In accordance vith Speci cation 1.0.rclicf valve shall be manually o ened unt ermo coup es ac tic monitors do earn of valv indicat steam is f loving from thc valve.AMENbMENT NO.2 29 P GE~

5 ci4ica4iom 3.5.(AU 0 1989 3.5.4~~~s&e Whenever the core spray systems, LPCI, HPCI, or RCIC required to be OPERABLE, the disc pipi from the pump disch ge of th e systems to the las block lve shall be filled.~a S g 3.C'.l.1 The folloving surveillance requirements shall bc adhered to assure that the discharge piping of the core spr systems, LPCI, HPCI, d RC C are filled: g.A 7 The sucti of the IC an pumps shal be aligned to the condensate s orage tank, and thc pressure ppression chamber head tank shall ormally be aligned to serve the dis arge piping of th RHR and CS pum s.The con sate head t may be used to s e the RHR and S discharge pipin if thc PSC head tank is unav lable.The pr sure indicato on the discha e of the RHR and CS umps shall in cate not less th listed belov.Pl-75-20 48 psig Pl-75-48 8 psig P1<<74-51 48 psig Pl-74-65 48 psig""<~~~<bio gr Z4w-~$04 gag IgyS 1.Every month t e RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems sha 1 e volte rom t hig po~t and vat flov detc ined 443 2.Fo ov ng any period vhere t LPCI or core spray systems hav not been r uired to be OPE LE, the di harge, iping of th inoperable stem hall be vent from the igh point prior to the return of the system to service.LA9'.Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage tank, thc discharge piping of the HPCI and RC s c ve e rom the gh poin f thc s stem had vater flov obse d on a monthly basis.4.When the RHRS and the CSS are r uired to bc PERAB the pre ure indicat vhic monit thc dischar lin shall be monitored daily and thc prcssure recorded.OBFH Unit 2 3.5/4.5-17 AMENDMENT tltO.I 6 9 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 8 CO COO~z.i MQV 22 1988 4.5 MIXER 00 G kppli.es the opcrationa status of e core and containmcnt cooling systems.lies to the surve lance rcq rements of thc co e and conta ent cooling sys when thc cor cspondipg limiti condi-tion for peration is in e feet.To assure the 0 LITT of the core and containm t cooling systems under all cond iona for which this cooling capab lity is an essential response to ant abnormalities. To veri the 0 ILITY of thc core and conta t cooling systems under all onditions for which this cooling pability is essential response to plant abnormalities. t aco 3 5'-((1)PRIOR TO STARTUP from a COLD COHDITIOK, or Ay l'cab4)(2)when there is irradiate fuel in the vessel and when the reactor vessel pressure is greater than atmospheric pressure, except aa specified in Specification 3.5.k.2.1.The CSS shall be OPKHBLE: SR s.s.).g Pn~sed]u.>+see.s..ao b.Rg sg p.g,>,L c+Simulated Automatic Actuation teat Pump OPBM-BILITX 9 Once/Q~Q&44LC441g 4p6kc Pcr Specifi-cation 1.0.c.~otor 0 eratcd V ve ILIA r Spe ifi-ca on 1 O.MM l.Core Spray System Testing.sZZc.l.(d.System flow rate: Each loop shall deliver at least 6250 gpm against a system head corres-ponding to a Once/&8Z m~?$BFH Unit 3 3.5/4.5-1 PAGE NENDMENTNO. t>O Ci +F1 f 105 psi'diffcreatial 3 5 J~(prcssure betvcea the reactor vehsel and the primary containment. 6c in hloge3 ZH I&bras 3.If Specification 3.5 h.l or Specification 3.5.k.cannot be met, the reactor shal be placed in the COLD SHOTDOWE COHDITI05 vithin hours.3C, L2, PC4'o nJ 8+H 2.If one CSS loop is inoperable, the reactor may remain in operation for a period aot to exceed 7 da rov ng 1 active components in iqC+ie~H the other CSS loop and th RHR stem CI mod e dicsc generator are OPERABLE.e.stable Per Che Valve Qpecifi tioa I.O.NN 5 3.5-.>Verify that each valve (manual y povcr-operatcd, or Bs automatic) in the inJection flovpath that is not locked, sealed, or othcr-vise secured in position, is in its oottsotg-Qyl position.additi~sotssiLLanos is ired.5cc~~gg;ecch'oit Qc change~r Se+ISV'S r,S, I A3 4.en the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERABLE pump and associated diesel generator shall bc OPERhBIZ, except vith the reactor vessel head remove as s ecificd in 3 h.r a specified ia 3.5.h.l Except that an automat c valve capable o autom ic r turn to its EC posit on vh aa ECCS signa is prcscn may bc in a position for another mode Scc gusttg~fio~+chcln)cf rr BAN ISTIC 3.5.2 BFH Unit 3 3,5/4.5-2 g,l Once/iB,s~aking Qcc1s-qua I~k Simulated Automatic Actuation Test RHRS shall be OPERABLE 8.The Q2.PRIOR TO STARTUP from a COLD CONDITION; or Applicab Iig (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.$.2 through 3.5.B 7.s83.5, l.0 Sg 3.S.I.Z oR PC~:~y~mme~~St~keff valve i~g~LPC(crust g,'t ts c4scd~C 3.S.l.Z Nsa Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation'or shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5-4 BFN Unit 3&o b.c~e.Pump OPERA-BILI~er Specification 1.0.MM Motor Opera-Per ted valve Specification OPERABILITY 1.0.MM Pump Flow Once/W J Rate men ths-4ag Testable Per Check Specification Va,lve 1.MM Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion,'n its 7 correc position.Once/Verify LPCI Once/subsystem cross-tie valve is closed~power removed from valve operator.Except that an auto tic valve cap le of auto-ma ic retu to ts CS posi on en an ECCS igna is present may be in a position for another mode of operation. NENOMEMT HO.I 77 i 0 skci4 ca~n a.s.SURVZnumCE REqUIZZmm UB 02 2~3~+5'o 4 With the reactor csscl prc sure less t 105 psig, th RHRS may b removed f om service except tvo pumps-c tainmcnt ooling mode and a ociated h at exchanger must r OPEiULB for a pc od not to cxc d 24 hour vhilc being drained o supp cssion ber quali y va r and fi ed vith p ry coo t quali vatcr prov ded that d ing cooldovn o loops v th one pump pe loop or o loop vith pumps, associ tcd diesc generators, in e core spray system a e OPBRhBLE.1.If ne RHR um (I mode)is inoperable, the reactor may remain in operation for a period not to excccd 7 days prov e e rema pumps (LPCI mode)and bo access paths of the LPCI mode and the S and in a ors rema OPE~~~~R Each LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Two LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system prcssure of 250 psig.2.hn air test on the drywell and tyne headers and nozzles shall bc conducted once/5 years.k vatcr test may be performed on the torus header in es See Zff5t)fi'Gabon &I Chases~c SP'iV (5 Tg g,g, g.q Sea$~,1'C'cJi gyral tsrS 3 f.I 4.If any'RHR um (LPCI mode ecome inoperable, the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.3C Bc,;r e&e3, in Izhrs BFH Unit 3 3.5/4.5-5 pAGE 5 oF~~hMENGMBlTNO. t a O Cl nt ent 8.Ron5 8+'If Specifications 3.5.B.l through 3.5.B.7 are not met,'a44keted d the reactor shall be placed in the COLD SHUTDOWN CONDITION within hours.3 R in~<3'~IPhrs A1 Sec JuSt 4>'capon Per Changes gr gP~ts rs s.s.~When e reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolagt injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specif ication 1.0.MM.10.If the conditions of Specification 3.5.A.5 are met, LPCI and containment coolin are not requi d 10.No additional surveillance required.BFN Unit 3 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 The B and D RHR p'umps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0;MM when the cross-connect capability is required.AMENDMENT NQ.g 77 pgsp~oF~t~ 8 AUG 02 t989 on 12.If one RHR pump or associated heat exchanger located on the unit cross-connection in unit 2 is i perable for any reason (eluding valve inopcrabilit pipe eak, etc.), the rc tor ma remain in operation for a eriod not to excee 30 days rovided thc remai RHR pump d associated diese generator a OPERABLE.3~14.I cross-co ction flov or heat emoval capabi ty is lost, thc un t may remain i peration for a pe od not to exce 10 days unless such capability is restored.h rec rculat n p ischarg valves shal b OPERAB PRIOR 0 ST (o closed f pe tted e cvhere in t se spe ficatio).13.Ho additional surveillance required.SR XS.14.All recirculation pump discharge valves shall be tested for OPERABIL during any period of gg 3~t~COLD SHUTDOWR COHDITIOH exceeding 48 hours, if OPERABILITY tests have not been performed during thc preceding 31 days.BFH Unit 3 3.5/4.5-8 AMENDMENT HO.I 4 0 PAG<~ S Stoic I g(ggt'EB i 4 1995 Sc<3'us,H f-:<chion Qe, ckclwy g I ST 5 Section 3'g (~ca>u~S'6;c tt'nnA, 0 ng c5 g~(et p~l STS Z87 4.9.A.4.(Cont'd)c.The loss of voltage and degraded voltage relays which start the diesel generators from the 4-kV shutdown boards shall be calibrated annually for trip and reset and thc measurements logged.These relays shall be calibrated as specified in able 4.9.A.4.c. d.4-kV shutdown board voltages shall bc recorded once every 12 hours.a.Accident signal logic system is OPERABLE.b.480-volt load s e logic system is OPERABLE.>ee 3<5(inca fjin far~go~'~(sTS S.g.(5.4-V V pard 3Rk2E a.Once sC r.s.i.u.c automatic transfer feature for 480-V RMOV boards 3D and 3E shall bc functionally tested to verify auto-transfer capability. 6.There shall bc a minimum of 35,280 gallons of diesel fuel in each of the 7-day diesel-gcncrator fuel tank assemblics. 5ea$wy;p;a.4on Qr~~Ri pFg BPS Qnit 3 3~9/4~9-7 AIItEHDMEÃf Rtt.I 8 9 PAGE 8'oF~~ 8 , NOV 0 4 1991 e.Loss of voltage and degraded voltage relays OPERhBLE on 4-kV shutdown boards 3Eh, 3EB, 3EC, and 3ED.f.Thc 480-V diesel auxiliary boards 3Eh and 3EB are encrgizcd. g.The 480-V reactor tor-opera d valve (V)boards 3D Ec 3E arc ergized th motor enerator)sets 3D 3Dh, 3ES, and 3Eh in service.Sec.s~<<"~~L'4~)~)4-~Be<(sr>s.s.7 4.Thc 250-V shutdown board 3EB battery, all three unit battcriea, a battery!charger for each battery, and associated battery boards arc OPERhBLE.Sc4>~ski((ek,oe fur C44 pg cfor OP~(pre g~q 4.a.(Delctcd)b.Once every 18 months,)the conditions under which the loss of voltage and degraded voltage relays are regni,rcd shall be simulated with an undervoltage on each shutdown board to deaoastrate that thc associated diesel generator will start.><3uStigi'agon fir Cfu~eS~SPu IST5 s BPS Unit 3 3.9/4.9W pAGE~AMENlpjp~TN 1 5 8

PERA S S.SPe.c.,g;~I'on Z.5.l 1%lOV 18 1988 10.When one 480-V shutdown board is found to be inoperable, the reactor vill be placed in HOT STANDBY CONDITION vithin 12 hours and COLD SHUTDOWH OHDITIOH within 24 hours.WStjCicah'on 4r Changing~r BAN I$75 P.g,-7 11.If on 480-V RMOV board g set s inoperab e, REA OR PO OPERATIO may co inue for perio not to exceed se en days p ovided th remain ng 80-V RMOV board sets and their associa d load remain 0 ERABLE.12 If any tvo 480-RMOV board sets ecome inop abLe, t e react r sha be pla ed in t e CO SHUTDO CONDI OH vithin 24 h urs.13.If t e r cerements or operation in the conditions specified by 3.9.B.1 through 3.9.B.12 cannot be met, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWH CONDITION vithin 24 hours.SeC g>~>.;~Hon 4.Cg~~CS+~ON I ST~SecAen Xg BFN Unit 3'MEtf0t,1ENT M.y p g PAGE~OF 11 FEB P 7 t991 3~5~4.5.D u 1.The equipme area cooler associated vi h each RHR ump and the e uipment a ea cooler ass ciated vi each set of core spra pumps (A an C or B d D)must b OPERAB at all tim vhen the ump or pum served by hat specif cooler is c sidered t be OPERABLE'. E h equipment rea cooler is crated in c unction vith e equipment erved by that articular c oler;therefore the equipm t area coole are teste at the same fre ency as th umps vhich th serve.2.en an equipme area coo er is not OPE LE, the p p(s)served y tha cooler ust be consi ere inoperabl for Technical Specificati purposes.C Co I ky);cubi l$)Prop seA Alod Pr SR E.g.l,8 The system shall be OPERAB enever there is irra ated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWH CONDITION or as specified in pecification 3.5.E.2.OPERABILITY shall be deter-mined within 12 hours a te reactor steam pressure 4 oM reaches 150 psig from a OLD CONDITION, r a t ely PR R TO ST TUP b usi an aux iary ste sup y.~S~.~r a.PropoSed hk K 4r SR.g.g.l.HPCI Subsystem testing shall be performed as fo vs++al or Simulated Once/18 Automatic months Actuation Test b.Pump 5R~~t.g OPERA-BILITY Per Specification.O.MM c.otor per-e a ed Va ve S cif cation OP RABIL Y 1.~]5/P 3.5.I.1 Prcfo+g Qpg d.Flow Rate at Once/W no a rect r 3 v s 1 kz.op a ng pr s re qco H l~/o.BFN Unit 3 PlO 3.5/4.5-1 AMENDMENT tl0 152 ls 0 on 3 So t FEg 0 7)99t~~R z.s.l,g e.Flow Rate at Once/18~psig months Ps g The HPCI pump shall deliver.at least 5000 gpm during, each flow rate test.l.'hov l4 5R p.g.(.~Verify that Once/Wm~each valve (manual, power-operated, or automatic) in the infection flow-path that is not locked, sealed, or otherwise secured in position, is in its correc osition.A7 2~tfo+std QCTTo&3~PtChor15 G-0 H F.R If the HPCI s stem is inoperable, t e reactor may remain in ope ation f period not to exceed'ys, provided the RHR LPCI OPE LE ICS are er'.Ac z~cd iR Ic 1't If Specifications 3.5.E.1 or 3.5.E.2 are not met, the reactor vessel pressur shall be reduced to 150'<<s psig or less within 24-3g ours.L2.Co e solation Cooli t RC CS cept hat an automatic va ve ca ble of utomatic ret rn to ts ECCS osition when an ECC signal s prese t may b in a positi for a other m de of opera ion.F.Reactor Core Isolation Coolin BFN Unit 3 The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY sha Sr 8-8 l5T$p.5$1.RCIC Subsystem testing shall be performed as follows: a.Simulated Auto-Once/18 matic Actuation months Test AMENDMENT NO.I 5 2 PAG~oF (1)PRIOR TO STARTUP from a COLD CONDITION, or,<PP~o&'li'Q 2./kh'ons Algol 3.ikhon g (2)whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 485 psig, except in the COLD DOWN CONDITION or as specified in 3.5.G.2 a 3.5.G.3 below.le 5 With one of the aSove required ADS valves inoperable, provided the H CI system, the core spray system and the LPCI syst are OPERABLE, s ore e inoperable ADS valve to OPERABLE status within 14 days or be in at least a HOT SHUTDOWN CONDITION within the next 12 hours and reduce reactor s earn dome ressure to Z psig within ho s.ISb g.3L With two or more of the above required ADS valves inoperable, be in at least a HOT SHUTDOWN CONDITION within 12 hours and reduce react r steam dome pressure to g~psig within hours.IW PC pO yA QI L5 R<oooor<d pioiion p,i pico 3.0.,3)t'ai t 4iA 7J)f5~DC'3 Lo>thin l3~~in: oi oooo~n 3 i gon<s~i.i> k 1.Six valves of the Automatic Depressurisation System shall be OPERABLE: 1.Durin each o crating cycl e following tests shall be performed on the ADS: tooaL Or gg p g1~oa.A simulated automatic actuation test shall be performed PRIOR TO STARTUP a ter~ch refute.ng outa e.nua surveys anc o the bpliefgval es is over+in g 4.6..2.ppapSC pie&~Sa 3.S.l l~.gi Pcafo5cJ QcQoA F BFN Unit 3 3.5/4.5-16 hNENOMENT NO.178..;.;;tx.,--(s EC 0'7 1994 3.6.C 4.6.C Z.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 2LZ'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for, thc air sampling system.The air sampling.system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout providing a temporary monitor.2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.>+c X~SWkimh on Q~Ch~6I ggp()5 fg y q 3.6.D.3.If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall be initiated aad the reactor shall be placed in thc COLD SHUTDOWN CONDITION vithin 24 hours.1.When more thaa one relief valve is kaovn to be failed, an orderly shutdovn shall bc initiated and the reactor depressurised to less than lOS psig vithin 24 hours The relief valves are not required to be OPERABLE in the COLD SHllTDOMN CQNDITIQN. 4.6.D.1.Approximately one-half of all relief valves shall be bench-checked or replaced vith a bcnchmhecked valve each operating cycle.hll 13 valves vill hav been checked or replaced upon the completion of every second cycle.<<<~r;~yon g~dtegCS 4<81-N l5T5 gag'R.g.]kr~rct jar fC."sR r.5.1.Ll ti t ermo oup e an a oust c mon tore do str am of hc ve ind cate stcam e lo ag f om the velvr 2.In accordance vith Specification 1.0.%5 c relief valve shall be manually o cned BPÃUnit 3 3.6/4.6-10 ANENMOrr NL I 88 GOOF

~Pec,'o~-S-t NY 9 l994 Whenever t core 8 ay systems, LPCI, HPCI, r RC required to be OPERAB the dis ar piping from'he ump disc rge these systems o the las bl ck valve shall filled.The s tion of the P pumps s all be aligned o the condensa storage tank, and t e press suppression amber hea tank s 1 normally b align d to se the dischar e piping f the and CS pump~e cond sate hea tank may be'us d to se e the aad CS dis harge px ing if th PSC head tank is unava able.Th pre88 e indica ors on th discha ge of the and CS umps shall i icate not less than listed b ow.1.Eve e RHRS CI and Conta nment Spray)and core spray systems, the dischar e pipin of these systems shall vened om/he h h pbjnt d wattr fl w%term ed.2.Following any period where the I or core spray sy tems have ot been equired o be OPERAB the scharge p ing of the ino erabl system 8 ll be vented f m the igh poin prior to the turn of e 8 8 tern't sly.s;/.1 The following surveillance requirements shall be adhered to assure that the discharge piping of the core spra systems, LPCI, HPCI, and RCIC are filled: Pl-75-0 Pl-75-4 Pl-74-51 Pl-74-65 48 p ig 48 ps 48 psig 48 psig 3~Whenever the HPCI R system is lined up to take suction from the condensate storage tank, the dischar e piping of the HPCI d RCI><Xsgg;mg.~Jr Chu@P~SFN ls T$p.g.p e vened from t high oint f the 8+tern and wate low observe on a monthly bas s.en the RHRS and the CS are r uired to b OPERABLE, t pre ure indicat s which monit the discha lines shall be onitored da y and he pressu e recorded.BFN Unit 3 3.5/4.5-17 NENOMENTRO, r 7 g q+Z'~QWI JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.Five current LCOs, 3.5.A, 3.5.B, 3.5.E, 3.5.G, and 3.5.H, have been combined into one proposed LCO (3.5.1).As such, the new LCO combines the three ECCS spray/injection Systems (HPCI, LPCI, and CS)into one LCO statement. The Bases continue to describe what components make up an ECCS subsystem. The new LCO statement also specifies that the six ADS valves are required.In addition, the ADS valve cycling requirements located in current Specification 4.6.D.1 are included as part of ADS operability. Thus, if an ADS valve does not cycle, the affected ECCS system is considered inoperable and the appropriate ACTION taken.A3 The Frequencies of"Once/operating cycle,""during each operating cycle," and"after each refueling outage" have been changed to"18'onths." This is considered equivalent since 18 months is the length of an operating cycle or a refueling outage cycle.The Frequencies of"Once/3 months" and"Per Specification I.O.HM" have been changed to"92 days," or"In accordance with the Inservice Testing program" as appropriate. The IST program test frequency for pumps is every 3 months and is currently defined by Specification I.O.MM.Therefore, this change is considered administrative in nature.The Frequency of"Once/month" has been changed to"31 days." 0 A4 Notes allowing actual vessel injection or ADS valve actuation to be excluded from this test (simulated automatic actuation test)have been added to proposed SR 3.5.1.9 and SR 3.5.1.10.Since the current BFN-UNITS 1, 2, 5 3 Revision 0 PAGE OF 0~I JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING I requirements state the test is"simulated" (i.e., valve actuation and vessel injection are inherently excluded), this allowance is considered administrative in nature.AS Proposed Condition H provides direction for various interrelationships between HPCI and ADS, and between LPCI and CS.The Action requires entry into LCO 3.0.3 for various combinations of inoperability which are consistent with the present required actions for the same various combinations. The actual requirements are not being changed.A6 The existing Applicability for Core Spray System (CSS)Operability (3.5.A.1), and Low Pressure Coolant Injection (LPCI)Operability (3.5.8.1), requires both systems to be Operable whenever irradiated fuel is in the vessel and prior to startup from a COLD CONDITION. The proposed change (LCO 3.5.1 Applicability) requires them to be Operable in Modes 1, 2 and 3.This change more clearly defines the conditions when CSS and LPCI are required to be Operable without changing the specific requirements which are currently located in individual specifications for each system.This change is, administrative because the same requirements for Operability currently listed in specific specifications will be labelled APPLICABILITY and apply to the entire ISTS Section 3.5.1, ECCS-Operating. The 3.5.A.2, 3.5.B.2, and 3.5.B.7 Applicabilities are only cross references and have been deleted.A7 The clarifying information contained in the"*" footnote has been moved to the proposed Bases for SR 3.5.1.2.The intent of the surveillance is to assure that the proper flow paths will exist for ECCS operation. The Bases clarifies that a valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.As such, moving this clarifying statement to the Bases is an administrative change.AS This requirement has been deleted since it only provides reference to another Specification, and does not provide any unique requirements. The format of the proposed BFN ISTS does not include providing"cross references." A9 Surveillance Requirements for HOV operability, and check valves that are required by the Inservice Testing (IST)Program, have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.BFN-UNITS 1, 2, 5 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING A10 The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system.Since the reactor steam dome pressure must be a 920 psig to perform SR 3.5.1.7 and a 150 psig to perform SR 3.5.1.8, sufficient time is allowed after adequate pressure is achieved to perform these tests.This is clarified by a Note in both SRs that state the Surveillances are not required to be performed until 12 hours after the specified reactor steam dome pressure is reached.CTS 3.5.E.1 already contains the context of the Note for the low pressure flow rate test.This is also consistent with interpretation of the current technical specification requirement for the high pressure flow rate test which is currently not modified by a Note.All The existing Applicabilities for High Pressure Coolant Injection (HPCI)Operability (3.5.E.1)and ADS (3.5.G.1)require the systems to be Operable whenever irradiated fuel is in the vessel and reactor pressure is greater than 150 psig (105 psig for ADS), except in the COLD SHUTDOWN CONDITION. The proposed change (LCO 3.5.1 Applicability) requires HPCI and ADS to be Operable in Modes 1, 2 and 3, except when reactor steam dome pressure is<150 psig.(Reference Justification L5 for the.change in applicability from<105 psig to<150 psig for ADS.)This change more clearly defines the conditions when HPCI and ADS are required to be Operable without changing the specific requirements which are currently located in the individual specifications. This change is administrative because the same requirements for Operability currently listed in the specific specifications will be labeled APPLICABILITY and apply to the entire ISTS Section 3.5.1,,ECCS-Operating. The 3.5.E.2, 3.5.G.2, and 3.5.G.3 Applicabilities are only cross references and have been deleted.A12 A finite Completion Time has been provided to verify RCIC OPERABILITY. The new.time is immediately and is considered administrative since this is an acceptable interpretation of the time to perform the current requirement. A13 CTS 3.9.A.3.h (for Unit 1 and 2)and 3.9.A.3.g (for Unit 3)require 480 V reactor motor operated valve (RMOV)boards to be energized with motor-generator (MG)sets in service.CTS 3.9.B.13 and 14 (for Unit 1 and 2)and ll and 12 (for Unit 3)provide Required Actions for when one or any two 480-V MG board sets become inoperable. There are two 480-V AC RMOV boards that contain MG sets in their feeder lines.The 480-V AC RMOV boards provide motive power to valves associated with the LPCI mode of the RHR system.The MG sets act as electrical isolators to prevent a BFN-UNITS 1, 2, 5 3 3 Revision 0 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING fault propagating between electrical divisions due to.an automatic transfer.Having an MG set out of service reduces the assurance that full RHR (LPCI)capacity will be available when required, therefore, the unit can only operate in this condition for 7 days.Having two MG sets out of service can considerably reduce equipment availability; therefore, the unit must be placed in Cold Shutdown within 24 hours.The'nability to provide power to the inboard injection valve and the recirculation pump discharge valve from either 4 kV board associated with an inoperable MG set would result in declaring the associated LPCI subsystems inoperable and entering the Actions required for LPCI.Since, the out of service times for LPCI and the MG sets are comparable, the deletion of the MG set actions is considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE Ml Proposed Action H requires LCO 3.0.3 be entered immediately which requires the plant to be in MODE 2 in 7 hours and MODE 3 within 13 hours when multiple ECCS subsystems are inoperable. This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.5.A.3, 3.5.B.4, 3.5.B.8, and 3.5.E.3).For CTS 3.5.G.2 it is slightly more restrictive since it requires the plant to be in MODE 2 in 7 hours where no action was required before.CTS require a shutdown to NODE 4 within 24 hours (except CTS 3.5.G.2 for ADS which also requires the plant be in NODE 3 in 12 hours)but does not stipulate how quickly MODE 3 must be reached.Reference Comment L12 which addresses the less restrictive change of being in NODE 3 in 13 hours versus 12 hours and NODE 4 (or<150 psig which is outside the applicability for ADS and HPCI)in 37 hours rather than 24 hours.Surveillance requirement SR 3.5.1.3 has been added to verify that ADS air supply header pressure is z 90 psig.This is a new Surveillance Requirement which verifies that sufficient air pressure exists in the ADS accumulators/receivers for reliable operation of ADS.Since this is a new Surveillance Requirement, it is an added restriction to plant operations. N3 With the reactor pressure<105 psig, CTS 3.5.B.2 allows the RHR System to be removed from service (except that two RHR pumps-containment cooling mode and associated heat exchangers must remain OPERABLE)for a period not to exceed 24 hours while being drained of suppression chamber quality water and filled with primary coolant quality water provided BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING that during cooldown two loops with one pump per loop or one loop with two pumps, and associated diesel generators, in the core spray system are OPERABLE.This appears to be an exception to CTS 3.5.A.2 8 3, which only allows one CSS loop (i.e., one loop with two pumps)to be inoperable for 7 days and an immediate shutdown if this cannot be met.The¹Note for 3.5.B.1 allows LPCI to be considered OPERABLE.during alignment and operation for shutdown cooling with reactor steam dome pressure<105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Proposed Specification 3.5.1 has a similar provision (Note to SR 3.5.1.2).Since the proposed Specification has no provision that would allow continued operation in MODE 3 with pressure<105 psig with two CS loops with one pump per loop OPERABLE, the proposed change is considered more restrictive. M4 An additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.5.A.3, 3.5.B.4, 3.5.B.S, and 3.5.E.3).CTS require a shutdown to MODE 4 within 24 hours but does not stipulate how quickly MODE 3 must be reached.Reference Comment L2 which addresses the less restrictive change of being in MODE 4 (or<150 psig for HPCI and ADS)in 36 hours rather than 24 hour s.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Not used.LA2 The details relating to system design and purpose have been relocated to the Bases.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the FSAR will be controlled by the provisions of 10 CFR 50.59.ECCS system operability determinations are described in the Bases.SR 3.5.1.1 will ensure maintenance of filled discharge piping.BFN-UNITS 1, 2,&3 Revision 0 PAGE jz JUSTIFICATION FOR CHANGES BFN ISTS 3.5-1-ECCS-OPERATING LA3 Details of the methods of performing surveillance test requirements and routine system status monitoring have been relocated to the Bases and procedures.'hanges to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA4 LA5 Any time the OPERABILITY of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. Therefore, explicit post maintenance Surveillance Requirements have been deleted from the Specifications. Also, proposed SR 3.0.1 and SR 3.0.4 require Surveillances to be current prior to declaring components operable.CTS 3.5.D/4.5.D, Equipment Area Coolers, are being relocated to plant procedures. Relocating requirements for the equipment area coolers does not preclude them from being maintained operable.They are required to be operable in order to support HPCI, RCIC, LPCI and CS system operability. If they become inoperable, the operability of the supported systems are required to be evaluated under the Safety Function Determination Program in Section 5.0 of the Technical Specifications. This change is consistent with NUREG-1433. LA6 CTS 3.5.E specifically states that HPCI Operability can be determined prior to startup by using an auxiliary steam supply in lieu of using reactor steam after reactor steam dome pressure reaches 150 psig.Details of the methods of performing this surveillance test requirement have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA7 CTS 4.5.H.l requires the discharge piping of RHR (LPCI and Containment Spray)to be vented from the high point and water level determined every month and prior to testing of these systems.The specific requirement to vent prior to testing has been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 8L 3 PAGE~OF~i;; 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING"Specific" Ll The phrase"actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying the ECCS subsystems/ADS actuate on an automatic initiation signal.This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill this requirement. Operability is adequately demonstrated in either case since the ECCS subsystems/ADS itself can not discriminate between"actual" or"simulated." L2 The time to reach MODE 4, Cold Shutdown (for LPCI and CS)and<150 psig (for HPCI and ADS)has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within, the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added for LPCI, CS and HPCI (Reference Comment M4 above).These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.A new Action (proposed ACTION 0)is being added to LCO 3.5.1 for the.condition of an inoperable HPCI System coincident with one inoperable low pressure ECCS injection/spray subsystem. The analysis summarized in the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996)demonstrates that adequate cooling is provided by the ADS system and the remaining operable low pressure injection/spray subsystems. However, the redundancy has been reduced such that another single failure may not maintain the ability to provide adequate core cooling.Therefore, an allowable outage time of 72 hours has been assigned to restore either the inoperable HPCI system or the inoperable low pressure injection/spray subsystem to operability. This change is consistent with NUREG-1433. L4 The allowable outage time for HPCI has been extended from 7 days to 14 days.Adequate core cooling can be provided by ADS and the low pressure ECCS subsystems. The 14 days is allowed only if all six ADS valves and the low pressure ECCS subsystems are operable.(The exception, LCO 3.5.1, Condition D, which allows operation for 72 hours with HPCI and one low pressure ECCS subsystem inoperable is addressed in Comment L3 above.)The 14 day Completion Time is based on the reliability study that evaluated the impact on ECCS availability (Memorandum from R.L.Baer (NRC)to V.Stello, Jr.(NRC),"Recommended Interim Revisions to BFN-UNITS 1, 2, 5.3 Revision 0 PAGE~GP~ 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING LCOs for ECCS Components," December 1, 1975).Factors contributing to the acceptability of allowing continued operations for 14 days with HPCI inoperable include: the similar functions of HPCI and RCIC, and that the RCIC is capable of performing the HPCI function, although at a substantially lower capacity;the continued availability of the full complement of ADS valves and the ADS System's capability in response to a small break LOCA;and, the continued availability of the full complement of low pressure ECCS subsystems which, in conjunction with ADS, are capable of responding to a small break LOCA.This change is consistent with NUREG-1433. L5~L6 The pressure at which ADS is required to be operable is increased to 150 psig to provide consistency of the operability requirements for.HPCI and RCIC.equipment. Small break loss of coolant accidents are not analyzed to occur at low pressures (i.e., between 105 and 150 psig).The ADS is required to operate to lower the pressure sufficiently so that the LPCI and CS systems can provide makeup to mitigate such accidents. Since these systems can begin to inject water into the reactor pressure vessel at pressures well above 150 psig, there is no safety significance in the ADS not being operable between 105 and 150 psig.\A new ACTION has been added (ACTION F), which allows an outage time of 72 hours when one ADS valve and a low pressure ECCS subsystem is inoperable. Currently, there is no allowed outage time when these two items are inoperable. The analysis summarized in the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996)demonstrates that adequate cooling is provided by the HPCI and the remaining operable low'pressure injection/spray system.However, the redundancy has been reduced such that another single failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Therefore, an allowable outage time of 72 hours has been assigned to restore either the inoperable ADS.valve or the inoperable low pressure injection/spray system.This change is consistent with NUREG-1433. L7 t out of service as BFN-UNITS 1, 2, 8.3 Revision 0 Current Technical Specifications only allow one LPCI pump to be inoperable. Proposed ACTION A allows two LPCI pumps, one per loop or two in one loop, to be inoperable for seven days.The BASES for ISTS 3.5.1 Required Action A.l state that the 7 day allowed outage time is justified because in this condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA.This justification is applicable for the LPCI function of RHR with one or two RHR (LPCI)pumps demonstrated by previous LOCA analyses performed for

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING BFN as well as the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996).Following postulated single failures, adequate core cooling can be provided by one loop of Core Spray (2 pumps)and two RHR (LPCI)pumps (either two pumps in one loop or one pump in two loops)in conjunction with HPCI and ADS.Therefore, this less restrictive change is acceptable based on the plant specific LOCA analysis perfqrmed for BFN.L8~L9 This change proposes to add a Note to current Surveillance Requirement 4.6.D.4 (proposed Surveillance Requirement 3.5.1,12)which states,"Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test." This change allows the Applicability of the Specification to be entered for 12 hours without performing the Surveillance Requirement. This allows for sufficient conditions to exist and allow the plant to stabilize within these conditions prior to performing the Surveillance. The normal outcome of the performance of a Surveillance is the successful completion which proves Operability. This change represents a relaxation over existing requirements. This change is consistent with NUREG-1433. Existing Surveillance Requirement 4.5.E.l.d requires verification that HPCI is capable of delivering at least 5000 gpm at normal reactor vessel operating pressure.The proposed surveillance, SR 3.5.1.7, requires verification of a minimum 5000 gpm HPCI flow rate with reactor pressure e 920 psig and<1010 psig.The HPCI performance test at high pressure is the second part of a two part test that verifies HPCI pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the HPCI pump is expected to perform.Performance of the HPCI test at both ends of the expected operating pressure range confirms that the HPCI pump and turbine are functioning in accordance with design specifications. The ability of the HPCI pump to perform at normal reactor vessel operating pressure has already been demonstrated. A small decrease in the pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. BFN-UNITS 1, 2, tIL 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING L10 Existing Surveillance Requirement 4.5.C.l.e requires verification that HPCI is capable of delivering at least 5000 gpm"at 150 psig reactor steam pressure." The proposed surveillance, SR 3.5.1.9, requires verification of a minimum 5000 gpm HPCI flow rate with reactor pressure at a 165 psig.This change is less restrictive because it could allow reactor operation at pressures up to 165 psig prior to performing the surveillance. Performance of HPCI pump testing draws steam from the reactor and could affect reactor pressure significantly. Therefore, HPCI pump testing must be performed when the Electro-Hydraulic Control (EHC)System for the main turbine is available and capable of regulating reactor pressure.Operating experience has demonstrated that reactor pressures as high as 165 psig may be required before the EHC system is capable of maintaining stable pressure during the performance of the HPCI test.The HPCI performance test at low pressure is the first part of a two part test that verifies HPCI pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the HPCI pump is expected to perform.Performance of the HPCI test at both ends of the expected operating pressure range confirms that the HPCI pump and turbine are functioning In accordance with design specifications. The ability of the HPCI pump to perform at the lowest required pressure of 150 psig has already been demonstrated. A small increase in the pressure at which the performance to design specifications is verified will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. Ll1 CTS 3.5.E.1 requires HPCI operability to be determined within 12 hours after reactor steam dome pressure reaches 150 psig from a COLD CONDITION. The proposed Note to SR 3.5.1.7 and 3.5.1.8 allows X2 hours to perform the test after reactor steam dome pressure and flow are adequate.This is based on the need to reach conditions appropriate for testing.The existing allowance to reach a given pressure only partially addresses the issue.This pressure can be attained, and with little or no steam flow, conditions would not be adequate to perform the test-potentially resulting in an undesired reactor depressurization. The proposed change recognizes the necessary conditions of steam flow and minimum pressure as well as a maximum pressure limitation and provides consistency of presentation of these conditions. The point in time during startup that testing would begin remains unchanged. The change simply changes when the 12 hour clock for performing the test 10 Revision 0 PAGED()p~ JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1>>ECCS-OPERATING must begin and permits testing to be completed in a reasonable period of time.L12 Proposed Condition H provides direction for various interrelationships between HPCI and ADS, and Between LPCI and CS.The Action requires entry into LCO 3.0.3 for various combinations of inoperability which are consistent with the present required actions for the same various combinations (CTS 3.5.A.3, 3.5.B.4, 3.5.B.8, and 3.5.E.3).However, the time to reach MODE 4, Cold Shutdown (for LPCI and CS)and<150 psig (for HPCI)has been extended from 24 hours to 37 hours and to reach MODE 3, Hot Shutdown (for ADS only)has been extended from 12 hours to 13 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 2 in 7 hours and MODE 3 (Hot Shutdown)within 13 hours has been added (Reference Comment Ml above).These times are consistent with the BMR Standard Technical Specifications, NUREG 1433.~L13 An alternate verification to ensure the LPCI cross tie between loops is isolated has been added for Unit 3.The addition of an alternate method of satisfying the surveillance requirement is considered less restrictive. Currently, the method used for all three units is to verify the LPCI cross tie is closed and power is removed from the valve operator.Unit 3 has a manual shutoff valve install between the cross tie for Loop I and Loop II.This verification ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other subsystem. Since the manual shutoff valve serves the same function as the power operated valve, the proposed change is considered acceptable. BFN-UNITS 1, 2,&3 Revision 0 PAGE~(PP (P S JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING RELOCATED SPECIFICATIONS Rl Browns Ferry Nuclear Plant consists of three units.The pump suction and heat exchanger discharge lines of one loop of RHR in Unit 1 (Loop II)are cross-connected to the pump suction and heat exchanger of Unit 2.Unit 2 and 3 systems are cross-connected in a similar manner.Technical Specification requirements related to RHR cross-tie capability between units have been deleted.The standby coolant supply connection and RHR crossties are provided to maintain long-term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR System associated with a given unit.They provide added long-term redundancy to the other ECC Systems and are designed to accommodate certain situations which, although unlikely to occur, could jeopardize the functioning of these systems.Neither the RHR cross-tie nor the standby coolant supply capability is assumed to function for mitigation of any transient or accident analyzed in the FSAR.Therefore, the operability requirements and surveillances associated with the cross-connection capability have been relocated to the Technical Requirements Manual (TRM).Changes to the TRM will be controlled in accordance with 10 CFR 50.59.0 BFN-UNITS 1, 2, 5 3 12 Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP ,4 AUG 02 1989 3.5.A Co 4.5.A Co S a S st SS 5'mS~C'~+on @c Cho,+45 Q(Qpg)5+5 p 4.5.A.l.d (Cont'd)105 psi dif'fcrential pressure betveen thc reactor vessel and the primary containment. e.Check Valve Per Specification 1.0.MM 2.If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days providing all active components in the other CSS loop and the RHR system (LPCI mode)and the diesel generators are OPERABLE.Once/Month f.Verify that each valve (manual, povcr-operated, or automatic) in the injection flovpath that is not locked, scaled, or other-visc sccurcd in position, is in its correct+position.3~If Specification 3.S.A.1 or Specification 3.5.A.2 cannot bc met, the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.2.Ho additional surveillance is required.4~Ql;caL: l.g LCo Ze 5.g When thc reactor vessel pressure is atmospheric and irradiated fuel is in the eactor vessel at least one core spray loop vith one OPERABLE um associated eccl generator shall be PERABLE except vit the reactor vcsscl head removed as specified in 3.5.A.5 r TO STARTUP as spccificd in 3.5.A.1.Except that an automatic valve capable of automati return to its ECCS positi vhcn an ECCS signal is present may be in a position for another mode of operation. Wc'ssFi(aH ~*a Clonic<4(BC'Sos l.S.J S~5<S4$e(a~~C~g~BPH 15'f5 34a2.BFH Unit 1 3'/4 5-2 FAGE~oF-7-hMENDMENT NO.16 9 4l Al 5fec;0;c)hoz r.s',z QEI: 15 f988 LCo~l cab'.Lsd Mhen irradiated fuel is in the reactor vessel and the zeac'tor vessel head is removed, core spray is not required to be OPERhBLE provided the cavity is flooded, the fuel pool gates are open and the fuel pool vater level is maintained above the lov level alarm point and rov one V ump ass ciate valv su ply thc andb coo ant s ply are OPERhBLE Profosc'L/tCT'I oW5'Its PssW SR 3 S'.Pisl'sos>~IF Z.5.z.S'r GSS Mhcn vork is in progress vhich has the potent al to drai the vessel, ual nitia on apabilit of ei er 1 SS L p or 1 pum vith he ca bility o injec ng va into he re cl assoc ate ese generator(s) are required.Me SusHk~m~on 0 r C~C 4o s~>mrs z,~.~BFK Unit 1 3.5/4.5-3 AMENDMENT 10.y 6 g kl 8.9.APQ'cab Jig Mo 3.5o2.Nag For SR xs;zA If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be~initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION ithin 24 hours.When the reactor vessel pressure is atmospheric and irradiated fuel is in the eactor vessel, at least one RHR loop with two pumps or two loops with one pump per loo shall be OPERABLE.c pumps soc ate mesc generators ust also be OPERABLE.prcssure coo an xngcction (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 8.No additional surveillance required.<r<rus+Amaon P r r h~W B<H t S'TS'R.s.l~R 9.When the reactor vessel pressure is atmospheric, the RHR pumps that are required to e OPERABLE shall be Z.demonstrated to be OPERABLE per Specification 1.0.MN.St'~3uSA/'cocoon Pg,r Cha~t Ac BPN ISIS g.g.a LCu Hept:ab;l+ If thc conditions of Specification 3.5.A.5 are met, LPCI and containment cooling arc not re uired.nce BFN Unit 1 When there is irradiated fuel in the reactor and thc reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MN when the cross-connect capability is re uired pre S~6C:~o ger Cha~S 4 BAN isTs z.s.('ANENOMENT NO.2 P 4 PAGE OF 4 -~i~e~eiiL%9KI~)i~i4ar aa4I~ri-4~ilier rt~f'lL I O'P PL J'O'I't O'I 9'1~l.~.~-~'~~~~.~II~~'~~~~II~~II~~~~~~~~I~~~~~~II~~~~~II~~~~~~~~~II~~~~~~~'I~~~~'~~~~I~~~'~~~~I~'~~~II~II~Cb~I~~~I.'t'I~~~~~~~~~~'~'~'~~~~~~~II~'~~~~~~:I'I~~~~~i~'~II~~~~~~~~~~~~~~II~~~~~~~'~II~~~~II'~~~~~II~II~~I~'~~~~~~~'~'I~II~~~~~~~

~<Wkstigic~C~g hsc B~I 5 3,ge R<<(+2 LIMITIHG COHDITIOHS FOR OPERATIOH sFecl+;cg,go& 7 s VEILLAHCE REQUIREMEHTS 3.7 4.7 cab Applies to the operating status of the primary and secondary containmcnt systems.Applies to thc primary and secondary containment integrity. OOQ~LvV To assure the integrity of the primary and secondary containment systems.To verify the integrity of the primary and secondary containment. A.C a S<3.S.~.Ag~t;~4 Leo@, g,g At any time that thc irradiated fuel is in th reactor vessel, an the nuc car s em s pressurized ov tmos hcric ressure or work is being done v has the potential to drain the vessel, thc pressure su 1 vatcr level d tern eratur s a e maintained vithin the folloving limits.a.Minimum water level~-6.25" (differential pressure control>0 paid)-7.25" (0 paid differen-tial pressure control)S'C Z,S,2>I ai Thc suppression chamber vater level be checked once cr enever heat s added to the suppression pool by testing of thc ECCS or relief valves the pool temperature shall be continually monitored and shall'be observed and logged every 5 minutes until the heat addition is terminated. Max~1N b imum vater level~BFH Unit 1 3.7/4.7-1 PAGE i 3..C.S utdo 4.9~G~0 S do Whenever the reactor is'n COLD SHUTDOWH COHDITIOH vith irradiated fuel in the reactor, the availability of electric power shall be as specified in Section.3.9.A except as specified herein.l.At least tvo units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be OPERABLE.l.Ho additional surveillance is required.See AsÃkcah~~N CQ yy 8~<lS'75 Sec+io~7,f 2.An additional source of pover energized and capable of supplying power to the units 1 and 2 shutdovn boards consisting of at least one of the following: a.One of the offsite pover sources specified in 3.9.A.1.c. b.A third OPERABLE diesel generator. 3.At least one 480-V shutdown board for each unit must be OPERABLE.4.One 480-V RMOV boar mg t is uire for ach boar (1D o 1E)ired t suppo t oper tion o the syst in ac ordanc vit 3.5.B.BFH Unit 1 3.9/4.9-15 AMENDMENT tttO, 2 0 3;-"..~~r'F

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF~

Cr.X'C.4;o 3.K.Q AUG OR 1998 3 5.h 4~5.h S SS Scc.awc$;P,~g,.8~~+~Sr's~8 s/4-5-h l.d (Cont'd)105 psi differential pressure betveen the reactor vessel and the primary containment. e.Check Valve Per Specification 1.0.MM 2.If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days providing all active components in the other CSS loop and the RHR system (LPCI mode)and the diesel generators are OPERhBLE.Once/Month f.Verify that each valve (manual, pover-operated, or automatic) in the~ection flovpath that is not locked, sealed'r other-vise secured in position, is in its correct+position.30 If Specification 3.5.h.l or Specification 3.5.1.2 cannot be met, the reactor shall be placed in the COLD SHUTDOWN COHDITIO hours.2.Ho additional surveillance is required.When the reactor vessel ressure is atmospheric and irradiated fuel is in the eactor vessel at least one core spray loop vith one OPERABLE pump assoc ated esel generator shall be PERhBLE except vith the reactor vessel head removed as s ecified in 3.5.h.5 r PRIOR TO as specified in 3.5.h.l.Except that an automatic valve capable of automatic return to its ECCS position" vhen an ECCS signal is present may be in a position for another mode of operation. c'st t 3 sagk i(i cd i~4~C~grg W Bf~1sT<3>I~cc ZNsJAi~,f'>a~ g~Ck~ggf+~~FIJ Isis z.g.2 BPH Unit 2 3.5/4'-2 hMENWENNO. 16 9 8 5 ~Pter fi~fio~3.5.~DEC 15 l988 Lco 3.S Z.When irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required to be OPERABLE provided the cavity is flooded, the fuel pool gates are open and the fuel pool vater level is maintained above the lov level alarm point and r v e one W puhy and ssociated alves g suppl the st dby coolant supply are OPERABLE.P~opocM Sg 8.<.B.A-'Popo~S~8.S:2.~CSS p<<posW ACnoaS.*When vork is in progress vhich ha the potenti l to drai the vess l, manual i tiation capab ity of eith l CSS Loop or RHR pump, ith the capabili of injecti te 411 the assoc ate ese enerator(s) are re uired s,~z~~$;4Lc 4'~4r 0~~~3.F.2-BFH Unit 2 3.5/4.5-3 AMENDMENT g6.~g 8 PAGE~i-iF~ S ent'nmen t 8.If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hour 8.No additional surveillance required.><~~s4i4c f~4, C~rri.4r Bf'N I s~g 9.4t't~W J.k>LCo 3.5.Q en the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.e pumps associated diesel generators must also RABLE Low pressure coolant injection (LPCI)may be, considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.HM.CQ O~~pl;~'.);t ~t e conditions o Specification 3.5.A.5 are met, LPCI and containment cooling re not re 404pk4%84~BFN Unit 2 When t ere is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours'~)3.5/4.5-7 The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.Sec.QNgf<f<~$ ~Qi (~~~~4 g~><sees B,g (AMENDMENT RD.2 2 3

~pc<Fice y i'o~Z.g~AUG 02 1989 IREMEHTS Whenever thc core spray systems, LPCI, HPCI, or RCIC are requi.red to be OPEBhBLE, the discharge piping from the pump discharge of these systems to thc last block valve'hall be filled.Thc suction of thc RCIC and HPCI pumps shall be aligned to the condensate storage tank, and the pressure suppression chamber head tank shall normally bc aligned to serve the discharge piping of the RHR and CS pumps.The condensatc head tank may be used to scrvc thc RHR and CS discharge piping if the PSC head tank is unavailable. The prcssure'ndicators on the discharge of the RHR and CS pumps shall indicate not less than listed belov.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 psig Pl-74>>65 48 psig Sec a4c44i~)o Q, PL QCt'r 85~Isrs 35/~353 The folloving surveillance requirements shall be adhered to assure that thc discharge'iping of the"core spray systems, LPCI , an RCI are filled.1.Every month an pr or toi e t t ng o the S (LPCI and Cont ent Spray)and co spra system the discharge p ping of these systems s a 3 e~te re e pqint and visitor flo~etermi ed 2~Folloving any per o v ere e,LPCI or co e spray systems ha e not been r ired~be OPE, the dis rge ping of th inoperable system shall be vent+from the high point prior to the return of thc system to 3.Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage tank, the discharge piping, of thc HPCI and RCIC shall be vented from the high point of the system and vater flov observed on a monthly basis.4.en the RHRS e SS are rc ired to e OPE LE, he pres re indi ators ich~p,)monito the di charge nes shall be onitored daily and the prcssure recorded.BFH Unit 2 3.5/4.5-17 AMStNENT Rt3.I 6 g PAGE

sec wgfgq,s:~P C4o~y gr gru I JMX S.g.2./f-~5 ceo ficRli~3.5,Q LIMZTIHG COHDITIOKS FOR OPE1RTI05 SURVEILLhHCE REQUIREMENTS ~7 4~7 C Applies to the operating status of the primary and secondary containment systems.Applies to the primary and secondary containment integrity. ~O~JJv To assure thc integrity of thc primary and secondary containment systems.Qhiaafze To verify the integrity of the primary and secondary containment. 1.At amr time that the irradiated fuel is in the reactor vessel and the nuclear system is pressurized above atmospheric prcssure or wor s e done whi has the potential to drain the vessel,'Ke prcssure suppression ool wats lcvc temperature ma nta ed within the following limits.a.Minianun water level~-6.25" (differential pressure control>0 paid',-7.25" (0 psid differen-tial prcssure control.SR as.z I a.The suppZession chamber water level b checked once er enever heat is added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall bc observed and logged every 5 minutes until the heat addition is terminated. b.Maximum water level~BFH Unit 2 3.7/4.7-1 OF V +CcifiC%4iue Z.S.Q VAN 0 8$99$3.9.C.0 o C d udo 4.AC 0 Co d S utdo whenever the reactor is in COLD SHUTDOWN COHDITIOH vith irradiated fuel in the reactor, the availability of electric pover shall be as specified in Section 3.9.A except as specified herein.~~ao aoclltlonsl surveillance is required.l.At least tvo Units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be OPERABLE.2.An additional source of pover energized and capable of supplying pover to the Units 1 and 2 shutdovn boards consisting of at least one of the folloving: Sea 3453 fscR'4d~Vol C lg-Jp!B~+I~~Seel~~9.f'.One of the offsite pover sources spec'fied in 3.9.A.l.c. b.A third OPERABLE diesel generator 3.At least one 480-V shutdown board for each unit must be OPERABLE.4.One 480-V RMOV board set required or each OV bo d (2D or require to supp t operatio of the RHR system accordance vith 3.5.B.9.BFH Unit 2 3.9/4.9-15 ANENOMENT RO.1 8 6 PAGE~OF~ 4 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP At deci gi earn AUB 02 lggg 4.5.k 4.5.k.l.d (Cont'd)GV&hkiWA on Ar C~(y g@g/pe g5yg 3.5.i 105 psi differential pressure between the reactor vessel and the yrimary contaimaent ~e.Testable Per Check Valve Specification Z.O.M 2.If one CSS loop is inoperable, thc reactor may remain in operatian for a period not to exceed 7 days providing all active components in thc other CSS loop and the RHR system (LPCI mode)and the dicscl generators are OPERkSLE.f.Verify that Onc each valve (manual y paver oyerated, or automatic) in the infection flowpath that is not locked, sealed, or othcr-vise secured in positiang is in its correct+position.3.If Specification 3.5.h.l or Syecificatian 3.5.k.2 cannot be met, the reactor shall be placed in the COLD SHOTDOWÃCOHDITIOE vithin 24 hours.2.No additional surveillance is re~ired.4.When the reactor vessel yressure is atmospheric and irradiated fuel is in the reactor vessel at least one (core syray laop vith one Z.S~CO umy associate esel generator shall b OPERABIZ exccyt vith the reactor vessel head rcmovcd as s ecificd in 3 AS.A.5 r RIOR as s ecificd in 3.5.k.l.Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may bc in a position for another mode of apcration. BFH Unit 3 5i-<~u&llca,giin Qp I~~A~xs 55 3.5;I 3.5/4'-2 See Z~sk:t';i,g~ 4 CIi-)~g~V Isis g.~.<PAGE~OF~NENONKNTNO. 1 O 0 Cl, Sfec.>gcW~ 3.5.2;DEC 15 1988*s.LGo gg.2 4'l'mb: lifp When irradiated fuel is in thc reactor vessel and the reactor vessel head ia removed, core spray is not required to be OPERABLE provided thc cavity is flooded, thc fuel pool gates are open and the fuel pool vater level is maintained above the lov evel alarm point ro ne p assoc atcd val cs s plying e stan y coo ant sup y are OPE LE Pw os'~s Sg, 3.g.g,g~cs~~oSect AC7 gong*When vork is in p gress vhich as thc po ential drain the ssel, man al init tion c ability o either CSS Loo or 1 pump, vi h the capa ility of a/ecting va cr into he reacto vessel c e generator(s) are required.Se<3'~f;~go~Q~gA BFH Unit 3 3.5/4.5-3 AMENOMENT N5.g P 2 Cl 8.9.LCn 3.sR X5 2,g If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop hall be OPERABLE.e pump assoc@a e xesel gener or must also be OPERABLE Low pressure coolant njection (LPCI)may be considered OPERABLE during alignment iand operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. Sc<5<sWVi ca,Ao n&<add tsrs~.s.~SR Z.5,2,S When the reactor vessel pressure is atmospheric, the RHR pumps that are required to e OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.~c 3'u5~'eaHo~ g I C4~b~e 1STS r,8.z 8.No additional surveillance required.~o 10 A(piiu b:[Q If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling~Le not required.e'FN Unit 3 en there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Becaus~cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The B and D RHR pumps on'nit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when, the cross-connect capability is required.e<gu~'Fi'WHon 4r~p&4/PE 15'f5 AMENDMENT No.X 77

'c NN 1 S 1994 Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled-5g X e fol owing surveillance requirements shall be adhered to assure that the discharge piping of the core s ray systems, LPCI, HPCI, an CIC are fille The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.Eve month and prio to t esting o the RS (L I and Con ament Spra and c re s ray s ems the dischar e piping of these systems shal e v e rom e xgh po t and wa flow dete~ned.2.o owing any period where the LPCI or core spray syst ve not been req red to OP LE the disch e i x 2 g P P g of th noperable sys shall be vente rom the high oint prior to the turn of the system to;service. Pl-75-20 Pl-75-48 Pl-74-51 PI-74-65 48 psig 48 psig 48 psig 48 psig 3.Whenever the HPCI or RCIC system is lined up to take suction from the condensate 'torage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis." 4.en the RS and he CSS are r uired to be OP BLE, the pre ure in cators ich moni o the d charge ines shall e onito ed dai and the pr sure rec rded.BPÃUnit 3 3.5/4.5-17 NENDMHfT i(0.1 78 PAGE 5 QF~

Skc>g~LDGTZBC CONDITIO?N ZOR OPERLTIOS SURVEILLAECE REQUIREHEHTS 3.7 4.7 kppliea to the operating'status of the priaary and secondary containNent ayateaa.hypliea to the primary and secondary contahunent integrity. To assure the integrity of the priaary and secondary contahaent ayateaa.To veri~the integrity of the priaLary and secondary cont ainccnt~gdd walib&~35'~l.Lt any togae that the irradiated fuel ia in the reactor vessel, e xmc e s ea yressurixed shore ataoapheric yreaaure or r e one~the yotential to drain the vessel, the pressure suppression pool water level tcRpera aa ta within the'following liaita.a.Madam water lerel~-6.25" (differential pressure control>0 yaid)-7.25" (0 ysid differen-tial yreaaure control)GAS.S.z.1 a.The auyyreaaion chaaber water level be checked once per glv$enerer heat a added to the suppression pool by testing, of the ECCS or relief valves the yool temperature shall be conthmally acmitored and shall be obaerred and logged every 5 minutes until the heat addition is terainated. .Maxima water level a See Q~g f'Non Q Chang)~~~'~<>34"Z.t+a BF5 Unit 3 3.7/4.7-1 PAGE~OF~ Cl 3.9.C.0~CO DIXIE S 0 4.9.C 0~CO)~~0 0 S DOWN Whenever the reactor is in the COLD SHUTDOWH COHDITIOH vith irradiated fuel in the reactor,.the availability of electric pover shall be as specified in Section 3.9.A except as specified herein.l.At least tvo Unit 3 diesel generators and their associated 4-kV shutdovn boards shall be OPERABLE.1.Ho additional surveillance is required.sc'c'5%f.'ca5~n far c~~+~><<JSVS 5 Ao q,~2.An additional source of pover energized and capable of supplying pover to the Unit 3 shutdown boards consisting of at least one of the folloving: a.One of the offsite pover sources specified in 3.9.A.l.c. b.A third OPERABLE diesel generator. 3.At least one Unit 3 480-V shutdown board must be OPERABLE.4.One 480-RNOV b ard motor nerator (mg)se is re uired fo each OV board (3D r 3E)r uired o suppo t operat on of e RHR system n accor nce v h 3.5.B.9~BFH Unit 3 3.9/4.9-14 AMENDMEQ S~g~z~8 s.VI 13USTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN ADMINISTRATIVE CHANGES A1'eformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.Surveillance Requirements for MOV operability that are required by the Inservice Testing (IST)Program have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.A3 CTS 3.9.C.4 requires one 480 V reactor motor operated valve (RMOV)board motor-generator (MG)set for each RMOV board required to support the RHR System in accordance with CTS 3.5.B.9.The 480-V AC RMOV boards provide motive power to valves associated with the LPCI mode of the RHR system.The MG sets act as electrical isolators to prevent a fault propagating between electrical divisions due to an automatic transfer.The inability to provide power to the inboard injection valve and the recirculation pump discharge valve from either 4 kV board associated with an inoperable MG set would result in declaring the associated LPCI subsystems inoperable and entering the Actions required for LPCI.Therefore, the deletion of the operability requirement associated with the MG sets in CTS 3.9.C.4 is considered administrative. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE/

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN TECHNICAL CHANGE-MORE RESTRICTIVE Ml Proposed ACTIONS A, B, C and 0 have been added to provide required actions be taken when LCO requirements can not be met.CTS 3.5.A.4 and 3.5.B.9 provide minimum requirements for ECCS subsystems when in MODE 4 and 5 (except with the spent fuel pool gates.removed and water level a the low level alarm setpoint of the spent fuel pool)but no action if these requirements are not met.Therefore, technical specifications are violated when these requirements can not be met and the default to TS 1.0.C.1 requires no action since the plant is already in Cold Shutdown.While from a compliance standpoint the proposed ACTIONS are less restrictive, from an operational perspective they are more restrictive since actions are required w'here there were none before.Proposed ACTION A allows 4 hours to restore a subsystem when only one of the required subsystems is inoperable and then proposed ACTION B requires action be initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs)immediately. The 4 hour Completion Time is considered acceptable based on engineering judgment that considers the remaining available subsystem and the low probability of a vessel draindown event during this period.With no required ECCS injection spray subsystems inoperable, proposed ACTION C requires action to be initiated immediately to suspend OPDRVs and at least one required subsystem be restored to OPERABLE status within 4 hours.If one subsystem can not be restored within four hours then Proposed ACTION D requires action be initiated immediately to restore secondary containment to OPERABLE status, to restore two standby gas treatment systems to OPERABLE status, and to restore isolation capability in each required secondary containment penetration flow path not isolated.These actions must be immediately initiated to minimize the probability of a vessel draindown and the subsequent potential for fission product release.0 M2 Proposed SR 3.5.2.1 has been added.SR 3.5.2.1 requires the suppression pool water be verified~a minimum level every 12 hours.CTS 3.7.A.1 (8 4.7.A.l.a)requires the suppression pool be verified e-6.25" with no differential pressure control once per day at any time irradiated fuel is in the reactor vessel, and the nuclear system is pressurized or work is being done which has the potential to drain the vessel.Therefore, proposed SR 3.5.2.1 is more restrictive since the frequency of performance has been increased from once per 24 hours to once per 12 hours.In addition, CTS only requires performance during atmospheric conditions when work is being done that has the potential to drain the vessel.Therefore, the proposed SR is more restrictive since it BFN-UNITS 1, 2, 5 3 Revision 0 PAULO';" 6 Cl JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN requires performance during MODES 4, and 5, except with the spent fuel storage pool gates removed and water greater than or equal to minimum level over the top of the reactor pressure vessel flange.The CTS requirement to check the maximum level during OPDRVs has not been included since Specification

3.5.2 concerns

the ability to maintain reactor water level using the suppression pool as a source of water.However, this level check is required for proposed Specifications 3.6.2.1 and 3.6.2.2 as it relates to Containment Systems.M3 Proposed SR 3.5.2.4, which requires a verification every 31 days that ECCS injection/spray valves are in their correct position, has been added.This provides assurance that the proper flow paths will exist for ECCS operation. This is more restrictive since BFN currently only requires this check during MODES 1, 2 and 3.M4 An SR has been added to require a system flow rate test for the Core Spray System during atmospheric conditions. While CTS (4.5.B.9)requires flow rate testing of the RHR pumps during atmospheric conditions as well as during MODES 1, 2, and 3, it only requires CSS flow rate testing during MODES 1, 2, and 3.The addition of this requirement is more restrictive. TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LA1 CTS 4.5.H.1 requires the discharge p'iping of RHR (LPCI and Containment Spray)to be vented from the high point and water level determined every month and prior to testing of these systems.The specific requirement to vent prior to testing has been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA2 Any time the OPERABILITY of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. Therefore, explicit post maintenance Surveillance Requirements have been deleted from the Specifications. Also, proposed SR 3.0.1 and SR 3.0.4 require Surveillances to be current prior to declaring components operable.PAGE~OF BFN-UNITS 1, 2,&3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN LA3 Details of the methods of performing surveillance test requirements and routine system status monitoring have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll CTS 3.5.A.5 requires manual initiation capability of either 1 CSS Loop or 1 RHR pump with capability of injecting water into the reactor vessel when work is in progress which has the potential to drain the vessel.The proposed Specification would not require the CSS or RHR (LPCI and containment cooling mode)system to be operable since LCO 3.5.2 applicability does not apply when the fuel pool gates are open and the fuel pool water level is maintained above the low level alarm setpoint.Therefore, the deletion of this requirement is considered less restrictive. The deletion is acceptable since the coolant inventory represented by this water level is sufficient to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. L2 The proposed LCO for ECCS-Shutdown is less restrictive since it only requires two low pressure ECCS subsystems to be OPERABLE.This can be fulfilled with any combination of RHR and CS subsystems. That is, two CS subsystems (a CS subsystem for Specification

3.5.2 consists

of at least one pump in one loop), two RHR subsystems (RHR subsystem for Specification

3.5.2 consists

of one pump in one loop), or one RHR subsystem and one CS subsystem OPERABLE.CTS 3.5.B.9 requires one RHR loop with two pumps or two RHR loops with one pumps per loop to be.OPERABLE.CTS 3.5.B.4 requires one CS loop with one pump per loop to be OPERABLE.Per CTS 3.5.A Bases the minimum requirement at atmospheric pressure is for one supply of makeup water to the core.Therefore, requiring two RHR pumps and one CS pump to be OPERABLE provides excess redundancy. In addition, since only one supply of makeup water is required, sufficient makeup water can be provided by two CS subsystems, two RHR subsystems, or one CS and one RHR subsystem. As such, the proposed Specification ensures redundancy by requiring any two low pressure ECCS subsystems to be OPERABLE.BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN RELOCATED SPECI F I CAT IONS Rl Browns Ferry Nuclear Plant consists of three units.The pump suction and heat exchanger discharge lines of one loop of RHR in Unit I (Loop II)are cross-connected to the pump suction and heat exchanger of Unit 2.Unit 2 and 3 systems are cross-connected in a similar mariner.Technical Specification requirements related to RHR cross-tie capability between units have been deleted.The standby coolant supply connection and RHR crossties are provided to maintain long-term'reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR System associated with a given unit.They provide added long-term redundancy to the other ECC Systems and are designed to accommodate certain situations which, although unlikely to occur, could jeopardize the functioning of these systems.Neither the RHR cross-tie nor the standby coolant supply capability is assumed to function for mitigation of any transient or accident analyzed in the FSAR.Therefore, the operability requirements and surveillances associated with the cross-connection capability have been relocated to the Technical Requirements Manual (TRM).Relocation to the TRM is in accordance with the"Application of Selection Criteria to BFN TS" and the NRC Final Policy Statement on Technical Specification Improvements. Refer to the application document discussion for additional information. BFN-UNITS I, 2,&3 Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF SfCC<CiCOQO m 3 i5 i3 FEB 0 7 199$3.5.E S st es ure Coo a t ectio C S 4.5.E essu e Coo a t S ste HPC S ect o 4.5.E.1 (Cont'd)Sce'uste C>>capon go>>ages Q>>gFv4 l5TS 3,5,(e.Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.f.Verify that Once/Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct*position.2.If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.*Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. eactor Co LCO P.S,>1.The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 5'g3,5g,g A'3 P~ppQ Vcr sR 3.5.3~RCIC Subsystem testing"hall be performed as follows: pc~i b C J a.Simulated Auto-Once/18 matic Actuati'on .-..oaths Test BFN Unit 1 3.5/4.5-14 AMENDMENT Ho.g 8 O=;:r'~P QF

5 ci+icgQoq Z 5'.3 NOV 24 1989 e determined vithin 12 hours after reactor steam pressure (oped reaches 150 psig from a COLD"~~~~COHDITIOH em veiny PQOQ T ST TU by ing Ran a~lory ste pl SP3.5.3.3 b.Pump OPERABILITY Per Specifi-cation 1.0.MM S R3.S,p,p PAyc se4 blue.Ai sRRs;p, c.M tor-0 era d er Va e eci PE BILI cat on 1.0.MM 9z~~T>d.Flov Rate at Once/N rma re ctor s ve sel pe ating pre ure Olo P 2~Qc7 le A IQ La 3~If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed W days if the HPCIS s OPERABLE during such time~ygi.,Q If Specifications .5.F.1 or 3.5.F.2 are not met, an~~~k.and the reactor shall be depressurized to less than 150 psig within 24'ours.ore~~~SRP.S>.q Se 3.S,q,p SR g.g,p g SR Z.S.'3.2 Be~Aodc 3 in IWhts Once/18 months f.Verify that Once each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc osition.A4, Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mod of o eration.e.Flov Rate at psig~ILS The RCIC pump shall deliver at least 600 gpm during each flov test.rlkqs BFH Unit 1 3.5/4.5-15 AMENOMENT NO.I 7 3 PAGE 8 >f'eciWi an 3,5,3 m 19 1994-Whee>~he core s ra s stems HPCI or ZC.ar required to OPERAB , t dis arge pipin from the pum disc rge the systems to e la t gg(b ck va ve shall be f lie Hm.5'g 8'.C.3.J The following surveillance requiremeats shall be adhered to assure that the discharge piping of the ore spray yetems, , HPCI, an RCIC a x ed: The ctioa of the RCIC d HPC pumps 11 aligned to the condeasa storage tank, d e pressure on chambe head tank shall normally be aligned to serve the discharge piping of the RHR aad CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC hea tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 peig 1-?4-65 48 psig Sr+Ycc+Q~,g~~<1ST5 3.g.)Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems shall be vented from the high point aad water flow determined. Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to serv 5 R 3.$.3.)3.Whenev RCIC system is lined up to take suction from the condensate stora e tank, the discharge iingof te RCIC s 1 e ven ed f m tge gh po t o the a'n at flo obs e cm s monthly as s.4.,When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.BFN Unit 1 3.5/4.5-17 NENOMENT NO, 2 06 PAGE~GP

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP~w>*~

(g~PC/i gi~%ion 3~3 FEB 0 7 1991.5.E s Co a t ct o 4.5.E S ste C S ssu Co t'ect o S stem C S 4.5.E.1 (Cont'd)e.Flov Rate at Once/18 150 psig months CL Dv$4limftow d'or<ha-P~s 4'o~BPhl ISIS g,s.i f.Verify that each valve (manual, pover-operated, or automatic) in thc injection flov-path that is not locked, sealed, or othervise secured in position, is in its correct+position.Once/Month The HPCI pump hhall deliver at least 5000 gpm during each flow rate test.2.If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSSo RHRS(LPCI), and RCICS arc OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdovn shall be initiated and the reactor vcsscl pressure shall be reduced to 150 psig or less vithin 24*Except that an automatic valve capable of automatic return to its ECCS position vhen an ECCS signal is present may be in a position for another mode of operati l C<1.Thc RCICS shall be OPERABLE vhenever there is irradiated fuel in the reactor vessel and the reactor vessel~ppi'c4.i~~prcssure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 1.RCIC Subsystem testing shall be performed as follovs: Pc or LI 5'R3.5.3.5 a.mulated Auto-Once/18 matic Actuation months Test P.,~~Noh r sa>5.>>BFN Unit 2>~sM ie 3.s3.9 3.5/4.5-14 NENOMENr No.190 PAGE N OF

S CRJXiCOJAJOM NOV 24 1999 Prop@~M4e~sg g.s.g f be determined vithin 12 hour after reactor steam pressure reaches 150 psig from a COLD CONDITION or a ernat ve y RIOR~0 RRRR~Oy naROR an auxilia~steam supply.'A3 4.5.F.1 (Cont d)SR'3.5.5.3 b.Pump OPERABILI TY Specifi-cation 1.0.MM tor-Operat d Per Va e Spec+i-OPE LITT ation~1.O.MM s~s.s.3.5 d.HJf Ol+O~4 4 sR 8.5.3*2JRRO S Flov Rate t orma react r v sel bgcra~in re ure Once ct Zo 4o IOIO PSJ+2.If the RCICS is inoperable, the reactor may remain in Ac TloQ operation for a period not to exceed days if thc HP is OPERABLE durin such time,.ppp f,R J'~J~4JJ P 3.If Specifications 3.5.F.1 or 3.5.F.2 are not met,~5aQA~k and the reactor shall be depressurized to ss than 150 psig vithin-hours.N,~cg.4 SR 3.S:~'f e.c.5 sf'.S.X,3 SA 3.S.3.9 5'~>53.Z f.Qp,i Flov Rate at sig~/45 Thc RCIC pump shall deliver at least 600 gpm during each flow test.Once/18 months Verify that nce/each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-visc secured in position, is in its correc po ition.Except that an automatic alvc capabl of automatic r urn to its ormal pos tion vhen a signal is prese may be in position for another mode of operation. BFH Unit 2 3.5/4.5-15 AMENOMENT Na.I 76~Cji' ~' 5')Ci'ccrc'u 3.5:3 UG 02 1989 he core s ray system PC HPCI or RCIC are requ red to be OPERABLE, the scharge p ing from he pump d charge of hest syst s to the st bloc valve sh 1 be fille Sg 8.S.>.e, folloving surveillance requirements shall be adhered to assure that the discharge piping of the core spra systems HPCI and CIC are filled: The suc on of th RCIC and HPCI pumps sha be alig ed to the condensate torage tank, an e pressure suppress on chamber head tank shall normally be aligne to serve the discharge piping of the RHR and CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate ot less than listed btlov.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 psig Pl-74-65 48 psi See X,q4<f'c~~io ~or C4 pg 4r 8P N ISTIC 2~g)gs a.)4.Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems shall be vented from the high point and vater flav determined. Folloving any period vhere the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the s stem ta servict Whenever the CI or RCIC system is lined up to take suction from the condensate storage tank, the discharge piping of the CI an RCIC e v ed fram t high po of the s tern and voter lov ob erve on a monthly basis.When t e an t e CSS are required to be OPERABLE, the pressure indicators vhich monitor the discharge lines shall be monitored daily and the pressure recorded.BFH Unit 2 3.5/4.5-17 AMEHbMENTHg. T 6 g PAGE~O~

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Cl ~<<~<'~ho B.S.3 FEB 0 7 l991 3.5.E essu e Coolant In ection S st PCIS 4.S.E i h ressure Coolant In ectio 4.5.E.1 (Cont'd)e.Flow Rate at Once/18 150 psig months 5 cd+iL5+jg'gg'on fia Ch+Qc5 Pnu BPn/]5y5 3 5.l Once/Month f.Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct*position.The HPCI pump shall deliver at least 5000 gpm during each flow rate test.2.If the HPCI system is inoperable, the"reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.*Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. Lco X5.3%Plica'ikj The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 1.RCIC Subsystem testing shall be performed a follows: A~or SP,g,g3.5 a.Simulated Auto-Once/18 matic Actuation months Test nap A h4w~sl s.S.~.s BFN Unit 3<>Rsed.~Q~Spy g 3 3.5/4.5-14 AMBDMEHTNO, yg p S HS FOR OPE s+~<;~+~-~NUV a+ious be determined vithin 12 hour t 4f'~~after reactor steam pressure reaches 150 psig from a COLD SC3.g,3.9 COHDITIOH r a t at e y PRI TO S RTUP usi an auxil st am sup ly.5'gg g p p b Pump OPERABILZ1Y .M tor-Operate Va e OPE ILITY er Specifi ation.O.MM r Sp cifi-cat n 1.0.Sg 3,g,pp d.rpo5rd go&sP.w,s;z.s SR Z.S.P.q l.6<CrS,a.>Sg3 Flov Rate at Once orma rea to v sel oper ting pr sur zo+o fo io t'so c)Flov Rate at Once/18 sig months C rgb The RCIC pump shall deliver at least 600 gpm during each flov test.2.If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed days if the Jq HPCIS is OPERABLE durin such time.ripe/'~md'atcl L.z 1 3.If Specifications 3.5.F.1 or 3.5.F.2 are not met, ee-D~~d-and the reactor shall be depressurized to less than 150 psig vithin 2A hours.3L>R'ua,'l H Slt'R 5.xz f.Br'n~3 ih f2hc g Verify that'ce/each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc sition.s*cept hat a aut atic v ve c able f aut matic re urn t its n rmal po tion en a igna is pre ent ma be in posi ion fo anoth r mo of operation. J~BFN Unit 3 3.5/4.5-15 AMENDMENT HO.1 4 4

~+4 Pmfion 3.5.3 NY i 9 894 WAmeovos-he core spray systems CI o C C are equired to b OPERAB E, t disc arge L.Al pipi from t e p disc arge of th se syst s to the la t block lve s ll be ille~RE.The following surveillance requirements shall be adhered to assure that the dis harge pipin of'he core spray systems, LPCI, HPCI, and RCIC are x e e sucti n of th R~C and HPC p s shal be ali ed to t e con sate s ra e t k the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.The condensate head teak may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.. The pressure indicators on the discharge of'he RHR and CS pumps shall indicate not lees than listed below.20 Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray systems, the discharge , piping of these systems shall be vented from the high point and water flow determined. Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system service.Pl-75-20 Pl-75-48 Pl-74-51 Pl-74-65 48 psig 48 psig 48 peig 48 psig S~RB.5.3 3 Whenever the PCI RCIC system is lined up to take suction from the condensate storage tank, the discharge pi iag of the CI an RCIC gee 3'uSb4ecaW< fpr Q)g,~~~For Sg'H 4 5g p e yea e rom g e bligh poin of Qe s nK wate low o serve on a monthly bas s.4.When e RHRS and the CSS ar required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.BFN Unit 3 3.5/4.5-17 NENOMENT tt0.I 7 B PAGE~OF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 The Frequency of"Once/month" has been changed to"31 days." The Frequencies of"Once/3 months" and"Per Specification 1.0NH" have been changed to"92 days." Since the proposed frequencies are equivalent, this change is considered administrative. A3 Notes allowing actual vessel injection to be excluded from this test (simulated automatic actuation test)have been added to proposed SR 3.5.3.5.Since the current requirements state the test is"simulated" (i.e., valve actuation and vessel injection are inherently excluded), this allowance is considered administrative in nature.A4 Surveillance Requirements for HOV operability that are required by the Inservice Testing Program have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.A5 The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system.Since the reactor steam dome pressure must be&20 psig to perform SR 3.5.3.3 and&50 psig to perform SR 3.5.3.4, sufficient time is allowed after adequate pressure is achieved to perform these tests.This is clarified by a Note in both SRs that state the Surveillances are not BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM required to be performed until 12 hours after the specified reactor steam dome pressure is reached.CTS 3.5.F.1 already contains the context of the Note for the low pressure flow rate test.This is also consistent with interpretation of the current technical specification requirement for the high pressure flow rate test which is currently not modified by a Note.A6 The clarifying information contained in the"*" footnote has been moved to the proposed Bases for SR 3.5.3.2.The intent of the surveillance is to assure that the proper flow paths will exist for RCIC System operation. The Bases clarifies that a valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.Moving this clarifying statement to the Bases is considered administrative in nature.A7~A8 A finite Completion Time has been provided to verify HPCI OPERABILITY. the new time is immediately and is considered administrative since this is an acceptable interpretation of the time to perform the current requirement. CTS 3.5.F.3 requires the reactor to be depressurized to less than 150 psig when CTS 3.5.F.1 and 2 cannot be met, while CTS 3.5.F.1 requires RCIC to be OPERABLE when reactor vessel pressure is above 150 psig.Proposed Required Action B.2 requires the vessel to be depressurized to x 150 psig.Since the intent of CTS is the same even though the CTS shutdown statement does not state"equal to," the addition of this requirement is considered administrative. TECHNICAL CHANGES-MORE RESTRICTIVE An additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specification (CTS 3.5.F.3).CTS require a shutdown to<150 psig within 24 hours but do not stipulate how quickly NODE 3 must be reached.Reference Comment L3 which addresses the less restrictive change of being~150 psig in 36 hours rather than 24 hours.BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl The details relating to system design and purpose have been relocated to the Bases.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the FSAR will be controlled by the provisions of 10 CFR 50.59.System operability determination, as described in the Bases and SR 3.5.3.1, will ensure maintenance of filled discharge piping.LA2 The details relating to methods of performing surveillance test requirements have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA3 CTS 3.5.F.1 specifically states that RCIC Operability can be determined prior to startup by using an auxiliary steam supply in lieu of using reactor steam after reactor steam dome pressure reaches 150 psig.Details of the methods of performing this surveillance test requirement have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll The phrase"actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying that the RCIC System actuates on an automatic initiation signal.This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adequately demonstrated in either case since the RCIC System itself can not discriminate between"actual" or"simulated." L2 BFN-UNITS 1, 2, 5 3 This change proposes to extend the current allowed outage time for the RCIC System from 7 days to 14 days.The 14 days are allowed only if the HPCI System is verified Operable immediately. Loss.of the RCIC System will not affect the overall plant capability to provide makeup inventory at high reactor pressure since the HPCI System is the only high pressure system assumed to function during a LOCA.However, the RCIC System is PAQE J pF 3~~Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM the preferred source of makeup for transients and certain abnormal events with no LOCA (RCIC as opposed to HPCI is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of the RPV water level).The 14 day completion time is also based on a reliability study that evaluated the impact on ECCS availability (Memorandum from R.L.Baer (NRC)to V.Stello, Jr.(NRC),"Recommended Interim Revisions to LCOs for ECCS Components," Oecember 1, 1975).Because of similar functions of HPCI and RCIC, and because HPCI is capable of performing the RCIC function, the allowed outage times determined for HPCI can be applied to RCIC.This change is consistent with NUREG-1433. L3 The time to reduce reactor steam dome pressure to a 150 psig has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added (See Comment Ml above).These times are consistent with the BMR Standard Technical Specifications, NUREG 1433.L4 Existing Surveillance Requirement 4.5.E.l.d requires verification that RCIC is capable of delivering at least 600 gpm at normal reactor vessel operating pressure.The proposed surveillance, SR 3.5.3.3, requires verification of a minimum 600 gpm RCIC flow rate with reactor pressure~920 psig and<1010 psig.The RCIC performance test at high pressure is the second part of a two part test that verifies RCIC pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the RCIC pump is expected to perform.Performance of the RCIC test at both ends of the expected operating pressure range confirms that the RCIC pump and turbine are functioning in accordance with design specifications. The ability of the RCIC pump to perform at normal reactor vessel operating pressure has already been demonstrated. A small decrease in the pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. L5 Existing Surveillance Requirement 4.5.F.l.e requires verification that RCIC is capable of delivering at least 600 gpm"at 150 psig reactor steam pressure." The proposed surveillance, SR 3.5.3.4', requires verification of a minimum 600 gpm RCIC flow rate with reactor pressure BFN-UNITS 1, 2, 5 3 Revision 0 ' JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM at 165 psig.This change is less restrictive because it could allow reactor operation at pressures up to 165 psig prior to performing the surveillance. Performance of RCIC pump testing draws steam from the reactor and could affect reactor pressure significantly. Therefore, RCIC pump testing must be performed when the Electro-Hydraulic Control (EHC)System for the main turbine is available and capable of.regulating reactor pressure.Operating experience has demonstrated that reactor pressures as high as 165 psig may be required before the EHC system is capable of maintaining stable pressure during the performance of the RCIC test.The RCIC performance test at low pressure is the first part of a two part test that verifies RCIC pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the RCIC pump is expected to perform.Performance of the RCIC test at both ends of the expected operating pressure range confirms that the RCIC pump and turbine are functioning In accordance with design specifications. The ability of the RCIC pump to perform at the lowest required pressure of 150 psig has already been demonstrated. A small increase in the pressure at which the performance to design specifications is verified will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. L6 CTS 3.5.F.1 requires operability to be determined within 12 hours after reactor steam dome pressure reaches 150 psig from a COLD CONDITION. The allowance for reactor steam dome pressure and flow to be adequate is based on the need to reach conditions appropriate for testing.The existing allowance to reach a given pressure only partially addresses the issue.This pressure can be attained, and with little or no steam flow, conditions would not be adequate to perform the test-potentially resulting in an undesired reactor depressurization. The proposed change recognizes the necessary conditions of steam flow and minimum pressure as well as a maximum pressure limitation and provides consistency of presentation of these conditions. The point in time during startup that testing would begin remains unchanged. The change simply changes when the 12 hour clock for performing the test must begin and permits testing to be completed in a reasonable period of time.BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5-ECCS AND RCIC SYSTEM BASES The Bases of the current Technical Specifications for this section (3.5.A, B, E, F, G, H, and 4.5)have been completely replaced by revised Bases that reflect the format and applicable content of proposed BFN-UNIT 1, 2,'and 3 ISTS Section 3.5, consistent with NUREG-1433. The revised Bases are as shown in the proposed BFN-UNIT 1, 2, and 3 Bases.BFN-UNITS 1,'2, 5 3 pAGF l Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP J S pdCiliCcyhon 7,C, (,/FEB 2 ames Ezo gy,/,]"/Pl<'~nb'ily erforming"open vessel" physics tests at pover levels not to exceed 5 MW(t).A b.Primary containment integrity is con rmed if e maxim allov le in egrated eakage rate, Lay does no't exceed the equi lent of pere t of the pr ry co ainmen volume r 24 h rs at the 49.6 psi design asis accident essure P'.Ce If 2 makeup to the rimary ontai ent ave aged ver 2 hours (correc ed f pr ssure, tempera ure, d ven ing op ations exc ds 542 CFH, it must b red ce to<42 SC vithin~ours<gioNpg or the reactor shall be emplaced in Hot Shu de RGTl&t 6)vithin the next 3k hours~2 2.a.Primary containment 4'/~~Pl'4 4aca~y shall be maintained at all times vhen the reactor is critical or vhen the reactor vater temperature is above 212 F and fuel is in the reactor vessel cep v e Primary co tainment n trogen consumpti n shall be monitor to etermin the averag dail nitrog cons tion or the ast 24 h urs.cessi leakage is ndicat d by a co umpti n rate f>I of e pr ry con ainm t free lume er 24 ours corre ed f dryv ll temper ture press re, venti op atio)at 49.6 psig Corr cted t no 1 d ell perati pressur of 1.psig, this value s 542 CFH.this value is exc eded, e action spec fied in 3.7.A.2.C shall be taken.5'R3-4~~~.<in accordance vith the Primary Containment Leakage Rate Testing Program.~4<<id'hu>~~n34ho, s BFK Unit 1 3.7/4.7-3 NENOMEg gg, pp8 FE822~i PAGE~OF 3.7/4.7W DF<c4'c'phon 3',(., l, 1 FEB 8 2 199j Pr,(.cour~ BFS Unit 1 3.7/4.7-5 hMENOMENT NO.2 2 8

Yush4caÃon gz PQ~y 0<8gaJ]Spy~~~<R 3.R.4~I'e I g.Perform required local leak rate tests, nc u ng t e r mary containment air lock leakage rate testing n accor ance v t t e Primary Containment Leakage Rate Testing Program.Eote: An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFH Unit 1 3.7/4.7-6 AMENDMENT NL?P.8 (1)Ef at ny time it is termined hat t criteri of 4.7.A.2.g's exceeded, repairs shall be initiated immediatel in Hob'in/s.0c~rs I~>><'/>n gg A~~pc.T(oN A (2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within~i hours.following detection of excessive local leakage, the reactor shall be until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by etest.The mann s earn one isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 3.7/4.7-8 When e prima conta ent's i ert the con inment s ll continuo ly monit red for gross eakage b viev of e inerti reg rements.This monit ing sys may be aken o t of se ice for nt e but s 1 be ret rned t service as soon pra icable.The in rior su aces the d Il and t ab e the evel one foot belo the n anal vat line out de surfac of th torus elov th vater ine 1 be suall ins cted ope sting cle or de riorat on signs st ctur e vith par cul attention to pip conn tions and suppo s and or signs of dist ess or displacement. BPS Uait 1 3.7/4.7-9 3.7.k.4 (Cont'4)4.7.4.4 (Coat'4)c.so Cryvell-suppression chamber vacuum breakers may'oe determined to be yerable for opening.Scc W Chewy>W B Ai~F5 3AI.i d<<<<f Spec<<<<cac<<oas 3~7.k<<<~aq 3.7.4.4.b, or 3.7.4.4.c. cannot be met, the aait shaLL be placed in a COLD SHUTDOWNÃCQ5D~ON in an orderly maaaer vithin 24 hours."-ach vacuum breaka valve shall be inspect~for yroyer oyeratioa oi="e valve aaC LM in accordance vith Spec'ficat'on 1.Q..C.Q~gi~le li 2-a tesc of the dr:ovel to suppression chambe st~care shall be conducted daring each o Lcc c Le Leaic rate Ls la~og.0.09 Lb/sec M yr~~containment acmosyhere vi" 1 ysi Cifie"mat'al 5.a.Containment atmosphere shall be reduced to Less than 4X oxygen vich aitrogea gas Curing reactor yover oyeration vith reactor cooLant yressare above 100 ysig, except as specified in 3~7.l<<5<<b.a.The yrimary.containment oxygea concentration shall be measured and recorde4 daily.The oxygea measurement shall be ad)usted to accoaat for the aacertainty of the metho4 used by adding a predetermined error faact'oa.b.Vichia the 24 hoar period subsequent to ylacing the reactor in th>>RN NDl foLloving a shat-Cova~the coatainmeat atmosphere oxygen coacentracioa shall be reduced to Less than 4%by volume and maintained Ln this condition. Deinerting may commence 24 hours prior to a shatCova~b.The methods used to measure the primary containment oxygen coaceatration shall be calibrated once every refueling cycle.c.If plaat control air ts being used to supply the yaeamatic coatrol system-inside primary coatainmcat, the reactor shall aot ba started, or Lf at yover, the reactor shall be brought to a COLD SHUTDOWN CQHDIT 05 vithia 24 hours.c.The coatroL air suyyly valve for the pneumatic coatroL systes Laside the pzmaz7 coatainmeat shall be verif'ea i closed prior to reactor star==and monthly thereafter. d.If Specificatioa 3.7.A.5.a aad 3.7.4.5.b cannot be met, aa orderly shutdova shall be initiated aa4 the reactor shall be in a COLD SHUTDOWN COHDITIOI vithia 24 hours.BFH Unit 1 3.7/4.7-11 AMEi fOMENT NQ, y g g PAGE I OF i UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP Cthe*Op, Qe J~J 2eae&0 3.k.i,J A p)i cg b i li 4 b.C~Alarm A Wlnpl 8 Primary containment maintained at all times when the reactor is critical or when the reactor water temperature is above 212 F and fuel is in the reactor vessel cept w er orming"open vessel" physics tests at power levels not to exceed 5 M(t).2 Primary conta nmen ntegri is confirme if t e max allo ble in egrate leakag rat , La, does n t excee th egu alent f 2 pe ent f the imary ontai t volum per 24 ours the 49.6 p ig desi basis ccident pressure, P.If 52 makeup to the primary con ainmen averag d over 24 urs (c recte for pres ure, t eratu e, and venti opera ons)ceeds 542.S, it uced to SC ithinA ours/or the reactor shall be placed in Hot Shutdown within the next hours Sf'rimary contai ent nitrogen consum tion s 1 be monit ed to d termine he aver e daily itrog cons ption f r the 1 st 24 ours.cessive eakage is ndicate by a H co umptio rate of>2X of t e prima contai ent free olume pe 24 hou (correct for d ell tempera re, pre sure, venting operati ns)at 49.6 p ig.Co rected to norma drywel opera ing pres re of.1 psi , this val is 54 SCFH.If this val e is ceded, e act on specified in 3.7.h.2.C shall be taken.~7.C.(.l e I er rm leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program d~hl Shu~a'~36 hOl4CS BFH Unit 2 3.7/4.7-3 hNENMENr NL 2 c 8

Sp<c jfirqQn Zi io.I>/FEB 8 259s BFR Unit 2 3.7/4.7W NENDMENT NC.2 4 3 PAGE~o..'I 5f ecsP:c'crgon 9.6, (.I FES 2 2 1996 4~~e t BFE Unit 2 3.7/4.7-5 hMENOMENT NO.2 4 3

FEB 22$9S~~~~A/3 3.4.lil g.Perform required local leak rate tests nc u ng e primary containment air lock leakage rate testi in accordance w e Primary Containment Leakage Rate Testing Program.See ruse<'~Son 4r c+~$Ar/go zsTs 3.o, I.~Hote: kn inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFH Unit 2 3.7/4.7-6 NENDMENT No.2 4 3 5

h: (1)If at any'me it is determi that e crite on of 4.7.2.g Is exceeded, repairs shall be initiated immediately. r~HojX-: g i~l2 N~r~+>>DC q;3g 4o~rg (2)CT(ofJ p,~fA p4 l AC7)o~B If conformance to the criterion of 4.7.A.2.g is not demonstrated within hour&f oil owing detection of excessive local leakage, the reactor shall b until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling'utage.If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. ~kQ MKgPJ ji (g)li~g~~>+>isrs g.S./g BFN Unit 2 3.7/4.7-8 PAGE j.o HmDar en the pr ry c tainment incr the ontainmena, shall be cont uously mo tored fo'r gr s leakage by reviev o the incr stem up r uircmen~This ear may bc taken ut of~rvice for maint a but be returned o serv ce as soon as pr ticable.e interior fac s of e dzpwell to ab e the leve one f t bel the no eater line outside surfac of the t belcnr eater 1 shall be sually pected ch opera iIlg le for de rioration and signs struc al e vlth parti ar atten ion to ping acti and s rts for signs o'f d tres or displac t BPI Unit 2 3.7/4.7-9

S'7 HOV 22 1888 5 acifica4iow E Co (\S 3.7.A.4 (Cont'd)4.7.A.4 (Cont d See v~dÃicJ~CQCs+AQ SAI isTS g.to I.7 c.Tvo dryvel'-suppression chamber vacuum breakers may be determined to be inopc abLe for opening.d.If Specifications 3.7.A.4.a.b, or.c cannot be met, thy uait shall be placed in a Cold Shutdovn condition in an orderly manner vithin 24 hours.SEE KcCSr tFiCArcrg~g CAAHQ~>R QFN<srs g.c,g~0 c.Each vacuum brcake" valve shall bc inspected'or proper operation of="e valve and limit switches in accordance with Specification L.O.HC.gp q.Q.).~d.A leak test of the d Jvell to suppression chamber tructure shall be conducted during each P l Accc table e rare~pg, 0.09 1 scc o pr mazy coatainmcat atmosphere with si differential a.Coatainmeat atmosphere shall bc reduced to less thaa 4X oxygen vith nitrogea gas during reactor pover operation Mich reactor.coolant pressure above 100/psig, except as specified ia 3.7.A.S.b. a.The primary coataiameat oxygen concentration shal'e measured and recorded daily.The oxygen measurement shall bc ad)usted to account"for the uncertainty of the method used by adding a predetermined error function.b.Mithin the 24-hour period subsequent to placing the reactor in the RUH mode folloving a shut-down, the containmeat atmosphere oxygen coaceatration shall bc reduced to less than 4X, by volume and maintained in this condition. Deinerting may commeace 24 hours prior to a shutdown.b.The methods used to measure the primary containment oxygen concentration shall be calibrated once every refueling cycle.c.If plant control air is being used to supply the pneumatic control system inside primary coatainment, the reactor shall aot be started, or if at pover, the reactor shall bc.brought to a Cold Shutdovn condition vithin 24 hours.c.The control air supply valve for the pneumatic control system inside the primary containment shall be verified closed prior to reactor startup and monthly thereafter. d.If Specification 3.7.A.S.a and 3.7.A.S.b cannot bc met, an orderly shutdovn shall be initiated and the reactor shall bc in a Cold Shutdown condition vithin 24 hours.BFH Unit 2 3.7/4.7-11 ph3c'gMg!T t'~C PAGE~OP ' UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP 5 Pc 4 i Ci cp on, g]2.a.Primary containment maintained at all times whea thc reactor is critical or when the reactor vatcr temperature is above 212 F and fuel is in the reactor vessel ccpt w i e pcrformiag"open vessel" physics tests at pover levels not to exceed MW t b.Primary ontaiamcnt i tcgrity s confi ed if th maxim allovab e in grated eakagc r te, La, does not exceed t e equi alent of 2 percen of the p imary co taiament volume er 24 h urs at t e 49.6 ps g design basis ccident P~c If H2 makeup to e primary c ntaiam t aver d over 24 hours (orrected or pr sure, t peratur and vent ng opera ions)cx eds 542 S , it t bc reduce to<5 SC vithin hours g or the reactor shall be c, placed in Hot Shutdovn'8 Swithin the next hours./2 Primary c taiament nitrogen coasumpt n shall monitor d to dete ine thc averag daily,ni rogen cons ption fo the las 24 urs.Ex essivc 1 ge is ndicate y a N2 c umptio rate of 2X of e prima conta ent frcc volume r 24 hou (corrc cd for ell'cmpe ature, p ssure, vcn ag opera ioas)a 49 psig.orrecte to n 1 d ell oper iag pressure f 1.1 ps g, this value is 542 SCFH If this value i excccd , thc action specifi in 3.7.k.2.c sha be taken.S 34~er leakage rate testing in accordance vith the Primary Contaiamcnt Leakage Rate Testing Program.~told Shukdoupn>n 34 goer g BFS Unit 3 3.7/4.7-3 lRNMNT NO.2 03 PAGE OF Sk<'4 0~Z<FEB 2 8 SSS BFR Unit 3 3.7/4.7W NIENDhfQT NQ, P 03 PA3E~OF~ eted BPH Unit 3 3.7/4.7-5 NBIDMEÃF RLP 0 3 PAGE~OF

g.Perform required local leak r eats nc u ng t e primary containment a r lo e rate testi in accordance v t t e Primary Containment Leakage Rate Testing Program.5ee 3~g40;rabin&<Ch~qcs&r SAN Xsrs Zt..l.<Rote: An inoperable air lock door does not invalidate the previou successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)@hen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFR Unit 3 3.7/4.7-6 NBIOMENT HO.2 03 PAaE h.(1)If at a time it is d ermined that the criterion o 4.7.A.2.g is exceeded, repairs shall be 3.nitiated immediately. pl~3 j~/2&Owed~Ho05 Q/~2C,A-<<(2)f conformance to the criterion of jhow'IorJ A 4.7.A.2.g is not demonstrated within~hour>following detection of excessive local eakage, the reactor shall be until+vta4 8 repairs are effected and the local leakage meets the acceptance criterion as demonstrated by etest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. 5j?t j~g jiCiGaf>ow 4~C44 p+AI BFN l>7-s g.z./.3 BFN Unit 3 3.7/4.7-8 PAGE

C t the pr ry con ainment i inerte thc ntainment hall contin usly mo ored for gros leakage eviev of e inert tea aike r remcnts.This monit rial sys may bc taken t of sc ce for maint e but shall be returned t service as soon as yrac icable.The interior surfaces of the dryvcll and torus above the level onc foot belov the normal vater line and outside surfaces of the torus belov the water linc shall be visually inspected each operating cycle for deterioration and any signs of structural d4RLgc vith particular attention to pipiag connections and.suyports and for signa of dis'tress or displacement. BPH Unit 3 3.7/4.7-9 0 3.7.A.4 (Cont'd)4.7.A.4 (Cont'd)c.Tvo dryvcll-suppression chamber vacuum breakers may be determined to be inoperable for opening.d.Sec'$45f&~og For chpggs+r 8cN fsrs 34(,7 If Specifications 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c, cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner vithin 24 hours.C~d.Once each operating cycle, each vacuum breaker valve shall be inspected for proper operation of the valve and limit svitches in accordance with Specification 1.0.MM.A leak test of thc dryvell to suppression chamber structure shall be conducted ur ng a o Acce table leak rate is.09 lb/sec o pr ary containment atmosphere vit 1 si differential. 5.0 5.0 a.Containment atmosphere shall be reduced to less than 4X, oxygen vith nitrogen gas during reactor paver operation vith reactor coolant pressure above 100/psig, except as specified in 3.7.A.5.b. a.The primary containmcnt oxygen concentration shall be measured and recorded daily.The oxygen measuremcnt shall be ad)usted to account for the uncertainty of the method used by adding a predetermined error function b..Within the 24-hour period subsequent to placing the reactor in the RUR mode folloving a shut-down, the containment atmosphere oxygen concentration shall bc reduced to less than 4X by volume and maintained in this condition. Deinerting may commence 24 hours prior to a shutdown.b.The methods used to measure the primary containmcnt oxygen concentration shall be calibrated once every rcfucling. cycle.c.If plant control air is being used to supply the pneumatic control system inside primary containmcnt, the reactor shall not be started, or if at pover, the reactor shall bc brought, to a Cold Shutdown condition vithin 24 hours.If the specifications of 3.7.A.5.a through 3.7.A.S.b cannot be mct, an orderly shutdown shall bc initiated and thc reactor shall bc in a Cold Shutdown condition ithin 24 hours.c.Thc control air supply valve for the pneumatic control system inside the primary containmcnt shall be verified closed prior to reactor sr.artup and monthly thereafter. Ce 3'u5hCs'cation fir<~g<f~BPu~5'r>Z.G.3.2.p~cE~oF 0 BFH Unit 3 3.7/4.7-11

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIMARY CONTAINMENT ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. A2 Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The definition of PRIMARY CONTAINMENT INTEGRITY has been deleted from the proposed Technical Specifications. In its place the requirement for primary containment is that it"shall be OPERABLE." This was done because of the confusion associated with these definitions compared to its use in the respective LCO.The change is editorial in that all the requirements are specifically addressed in the proposed LCO for the primary containment along with the remainder of the LCOs in the Containment Systems Primary Containment subsection (e.g., air locks, isolation valves, suppression pool).Therefore, the change is purely a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433.A3 CTS 4.7.A.2.k requirements for visual inspection of the drywell and torus surfaces are also contained in 10 CFR 50, Appendix J.These regulations require licensee compliance and cannot be revised by the licensee.These details of the regulations within CTS are repetitious and unnecessary. Therefore, the details also found in Appendix J have been deleted.This is considered a presentation preference and as such is considered an administrative change.BFN-UNITS 1, 2, 8L 3 Revision 0 PAGE~OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIHARY CONTAINHENT A4 A5 CTS 3.7.A.2.b provides acceptance. criteria for integrated leak rate testing, which is redundant to those contained in Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3)requirements. The definition of L.is provided in proposed BFN ISTS 1.1 and need not be repeated here.As such, this deletion is considered administrative. The acceptance criteria for the leak test of the drywell to suppression chamber structure has been changed from 0.09 lb/sec of primary containment atmosphere at 1 psid to 0.25 inches of water for 10 minutes.Since these values are equivalent this is considered an administrative change.A6 CTS 4.7.A.2.h(1) requires repairs to be initiated immediately when it is determined the criterion of 4.7.A.2.g is exceeded.CTS 4.7.A.2.g requires LLRTs to be performed in accordance with the Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3).CTS 4.7.A.2.h(2) then allows 48 hours to demonstrate 4.7.A.2.g can be met following detection of excessive local leakage.Since repairs are typically initiated immediately and proposed BFN ISTS ACTION A will only allow 1 hour"to restore primary containment to OPERABLE status prior to requir'ing the initiation of a shutdown (reference Justification H2 below), CTS 4.7.A.2.h(1) has been deleted.TECHNICAL CHANGES-NORE RESTRICTIVE Hl CTS 3.7.A.2.a requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.1 applicability is HODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in HODE 2, not critical and<212'F.H2 Proposed Action A is more restrictive than CTS 3.7.A.2.c since the time allowed to reduce excessive nitrogen leakage prior to initiating a shutdown has been reduced from 8 hours to 1 hour.The time allotted to place the unit in Hot Shutdown (HODE 3)has been reduced from 16 hours to 12 hours.Proposed Action B requires the unit to be placed in Cold Shutdown (HODE 4), whereas, CTS 3.7.A.2.c only requires the unit to be placed in Hot Shutdown.In addition, CTS 4.7.A.2.h.(2) allows 48 hours to demonstrate conformance to Appendix J following detection of excessive local leakage BFN-UNITS 1, 2, 5 3 Revision 0 PAGE R op 3

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIMARY CONTAINMENT and then requires a plant shutdown if conformance can not be demonstrated. CTS does not specify a completion time for shutdown and does not specify whether shutdown is to the Hot or Cold Shutdown Condition. The Proposed Actions A and B are more restrictive since they only allow 1 hour to restore primary containment and then require the unit be in MODE 3 in 12 and MODE 4 in 36 hours.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" programs.LCl The conti nuous leak rate monitor does not necessarily relate directly to primary containment operability. In general, the BWR Standard Technical Specifications, NUREG 1433, do not specify indication-only or alarm-only equipment to be OPERABLE to support operability of a system or component. Control of the availability of, and necessary compensatory activities if not available for, indications, monitoring instruments, and alarms are addressed by plant operational procedures and policies.Therefore, the continuous leak rate monitor, and associated alarm surveillances and actions will be relocated to a licensee controlled document.Any changes will require a 10 CFR 50.59 evaluation. LA1 The details relating to routine monitoring of plant status and operations parameters that reflect primary containment operability and the methods of performing this monitoring have been relocated to the Bases and procedures. Acceptance criteria for primary containment N, leakage (i.e., makeup consumption) have been relocated to procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled BFN-UNITS 1, 2,&3'AGE QP~Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 SAci 0i c~FEB 2 2 1996 At 2.a.Prima conta nment inte ity s ll be mai aincd t all imes vh the cactor is cr ical o vhen e re tor v ter cmpcr urc i abov 21'P and f el is n the ea or vcss cxc t wh e perf rmi"op vcs el" physics sts a po er levels not to exceed 5 t%(t).b.Primary containment integrity is confirmed if thc maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of.thc primary containment volume per 24 hours at thc 49.6 psig design basis accident pressure, Pa.c.If H2 makeup to the primary containment avcragcd over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must bc reduced to c 542 SCFH vithin 8 hours or thc reactor shall be placed in Hot Shutdovn vithin the next 16 hours.2.te ated Leak Rate estf Primary containment nitrogen consumption shall bc monitored to determine the average daily'itrogen consumption for the last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCPH.If this value is exceeded, the action specified in, 3.7.L.2.C shall be taken.Perform leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.gee guc+4'mb'+n Ar Chanyeg 4i BF~iSVS 3.C..i.l l-co p,5,1,2.8 lic4b.lip PI<RsR ACT'Iow5 A+g A]t'<opsy AJoH gg~/b;17oAS W3'Af$4 SR 3,g, t, f BFH Unit 1 3.7/4.7-3~AGE OF Sce XusHC'cab'on Qr@~ay'&<B~W tST5~~~~SR3.a.(.g. I g.Perfo required local leak rate tests including t e primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.Rote: An inoperable air lock 5g 3 g i g door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)when tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)when the air lock is pressurized to (g 2.5 psig for at least 15 minutes).<<e~~Wi'ca~n Qi C4~s 8FN)ST$5,g,I~BFH Unit 1 3'/4.7-6 AMENDMENT NQ.2 2 8 PAGE~QF g. Zf at any ime i" is det ined t at the iterio of 4.7.A.2.g exceeded, repairs shall be initiated immediately. ~~~3 I I24~/Hop/c/g g ggg within-48-ho rs f ollowing 2 4 HZ detection of excessive local leakage, the eactor shall be until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.(2)Ef conformance to the criterion of 4;7.A.2.g is not demonstrated The main steamline isolation valves shall be tested at a prgssure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 P.7/4.7-8 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

SPCCi4iCCrf)On "7 6-I-2 FE8 P, 2%6 2.a.Primary ontai ent integri y sha be mainta ned a all times vhen e re tor is critical I or v en th react vat r t cratu is a ve 2 2'F fuel s in e re ctor v esel cept ile pcrfo ng"op ves el" physic tests at po er level not to exceed b.Primary containment integrity is confirmed if the maximum allovable integrated leakage rate, La, docs not exceed the equivalent of 2 percent of thc primary containment volume per 24 hours at thc 49.6 peig design basis accident pressure, Pa.c.If S2 makeup to thc primary containment averaged over 24 hours (corrected for'rcssure, temperature, and venting operations) excccds 542 SCFH, it must be reduced to<542 SCFH vithin 8 hours or the reactor shall bc placed in Hot Shutdovn vithin the next 16 hours.2.Pr'imary containment nitrogen consumption shall be monitored to dctcrmine the average daily nitrogen consumption for thc last 24 hours.Excessive leakage is indicated by a E2 consumption rate of>2X of the primary containment free volume per 24 hours (corrcctcd for dryvell temperature, pressure, and venting operations) at 49.6 peig.Corrected to normal dryvell operating pressure of 1.1 peig, this value is 542 SCFH.If this value is exceeded, the action spccificd in 3.7.k.2.C shall be taken.Perform lcakagc rate testing in accordance vith thc Primary Containment Leakage Rate Testing Program.~<~~~iWWn 4<C~C5 A~8FhJ ISTIC ACO g, (o.t.2 Ppp))~g;)', P f'no@i~<RCT<o~a/I>b Pno sea 4u4 I tg p Acr<o+5]go g<J 5k'.6.lit 2 BFH Unit 2 3.7/4.7-3 hMENMNr N.2 c 8 PAGE R OP~ Ai I:ES 8 2896 5<c 5~~$icqgan Qr C/ggg5 Qr SPN l5i5 Z.6, l.l~~~~58 3.b.I~>~~g.Perform required local leak rate tests, includi the primary containment air lock leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.Hote: kn inoperable air lock door does not invalidate the previous successful performance , of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).Se'e Su5 tea'can'en Qr'kCi~gC'5 Ai 8<N l57<g.g.i~BFH Unit 2 3.7/4.7-6 NENDMENT N.243 pAGE X o~~ ~r w>>n a.>>>,/h.Ql)Zf at y time it i dete ined that the cri erion of 4.7.A.2.g i exceeded, epairs shall be initiated immediately. In~p~3~2 He>>rj p go~g~4 haul Rpgu>l cA Acfi'p<<C.2.P~pa~RCy<<>>Q Ac.f>~C.I P,c, ripe b (2)f'onformance to the criterion of 4.7.A.2.g is not d strated within ours following detection of excessive local leakage, the reactor shall b shet-down until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Ef the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. SeC Xug$,'f;~p~gpg hr 5/WIST~36 1..3 BFN Unit 2 3.7/4.7-8 PAGE~0~ 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 2.a.Prima ontainment int ity shal e ntained all times hen th eactor is itical or v the rea r vater t'rature above 2 F fuel in the actor vessel cept v e perf ing"o vessel" p sics tes at pover evels no to exceed%l(t)b.Primary containment integrity is confirmed if the maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.c.If H2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to<542 SCFH vithin 8 hours or the reactor shall be placed in Hot Shutdovn vithin the next 16 hours.2.I e rated Leak Rate Testi Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for the last 24 hours.Excessive leakage is indicated by a E2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal dryvell operating pressure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified in 3.7.k.2.c shall be taken.Perform leakage rate testing in a'ccordance vith the Prima Containment Leakage Rate Testing Program.5ee~~+;4i ca fjon4r Chang ts'~8@v 1ST'.C;I.I ~c.g.b, l.2 I;i~b:l&t'ro scd 4'TYPES 8 FB~>+<<g Sole Jyg y+'o<S%3 fno acA, Sk 3,k.l.l.BFS Unit 3 3.7/4.7-3 NBfDMNT go.P g 3 f'ACiE~QF~

5'cc~wSHAcafjon Qr PQ~<~8<m t Srs z.c,.i.]5 g.Perform equ red local le rate tests, includi the primary conta nment air lock leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.Hote: An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).sec WusA'C>cab'on 0)C~~6R di=nr g 7S g, g.i~BFR Unit 3 3.7/4.7-6 NBlDMST HO.2 03 pAGE 3 OF~ SFe C i Pica on p.(.I-2 h (1)at y ti it i dete ined hat th crit ion 4.A.2.g is exc ded, epair shal be in tiate immed atel 4euwcd l@hrn C.2.Aft'SC4 4Abn Co)(2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within hours following etection of cxcessivc local leakage, the reactor hall)trio v g repairs are i n o~3 effected and th OC Q I local leakage me the acceptance criterion as demonstrated by QZ retest.See 3iu+g;c~gn Foe ('-~gg)cg Qg, 8<<tsTs S.b.l.3<<~)K sc'cking The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall bc performed,to correct thc condition. BPR Unit 3 3.7/4.7-e PAGE~~"~

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.2-PRIMARY CONTAINMENT AIR LOCK ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore,-understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 CTS 4.7.A.2.h(1) requires repairs to be initiated immediately when it is determined the criterion of 4.7.A.2.g is exceeded.CTS 4.7.A.2.g requires LLRTs to be performed in accordance with the Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3).CTS 4.7.A.2.h(2) then allows 48 hours to demonstrate 4.7.A.2.g can be met following detection of excessive local leakage.Since repairs are typically initiated immediately and proposed Required Action C.1 for 3.6.1.2 requires action be initiated to evaluate the primary containment overall leakage rate using the current air lock results and ACTION A of ISTS 3.6.1.1 will only allow 1 hour to restore primary containment to OPERABLE status prior to requiring the initiation of a shutdown (reference Justification M2 for Specification 3.6.1.1), CTS 4.7.A.2.h(1) has been deleted.TECHNICAL CHANGES-MORE RESTRICTIVE The current requirements for the air lock are located within the primary containment TS requirements. The current definition of primary containment integrity requires only one air lock door to be closed and sealed (i.e., the seal mechanism intact and sealing the door).Thus, no actions are required if one door is inoperable provided the other door is OPERABLE, since primary containment integrity only requires the one door.The proposed LCO requires the entire air lock to be OPERABLE, BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES'FN ISTS 3.6.1.2-PRIHARY CONTAINHENT AIR LOCK which includes both doors, as well as the interlock mechanism and the leak-tightness of the barrel.ACTIONS are provided (proposed ACTIONS A and B)to ensure that if one door or its interlock mechanism is inoperable, the other door is closed, locked and periodically verified to be closed and locked.If the interlock mechanism is inoperable, an allowance is provided to open the door provided a dedicated individual controls the access.Notes are provided to allow, the locked closed'verification to be performed administratively if the door is in a limited access area.These two new actions are not applicable, however, if the entire air lock is inoperable (as stated in proposed Note 1 to both ACTIONS A and B).To ensure that the primary containment LCO will be entered if air lock leakage results in exceeding overall primary containment leakage, NOTE 2 to the ACTIONS is also included.Overall, these new ACTIONS provide additional restrictions to plant operation. CTS 4.7.A.2.h requires repairs to be initiated immediately when it is determined that the criterion of 4.7.A.2.g is exceeded and if conformance to these criterion is not demonstrated within 48 hours following detection of excessive local leakage, a reactor shutdown is required.ACTION C of the proposed Specification requires the licensee to initiate action to evaluate primary containment overall leakage rate using the current air lock test results immediately, verify an air lock door closed within 1 hour and restore the air lock to OPERABLE status within 24 hours.If required ACTION C and the associate Completion Time is not met, the unit must be in HODE 3 in 12 hours and HODE 4 in 36 hours.This is more restrictive than current requirements. This change adds a Surveillance to verify the interlock mechanism works properly (only one door can be opened at a time).This will ensure that one door is always closed which maintains containment integrity. The addition of new requirements represents a more restrictive change.The current requirements for the air lock are located within the primary containment TS requirements (CTS 3.7.A.2.a), which requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.2 applicability is HODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in HODE 2, not critical, and<212'F.BFN-UNITS 1, 2,&3 Revision 0 P@GF g QF UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP

4.7.A.a Co a 2oae m>Z.<l 3 R pp)scab'>I i b.C~Primary containment integrity shall be maintained at all times vhen thc reactor is critical or vhcn the reactor vater temperature is above 212 F and fuel ie in thc reactor veeec ing"open vessel" physics tests at pover levels not to exceed 5 HW(t).Primary containment integrity ie confirmed if the maximum allovablc integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.If E2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be rcduccd to<542 SCFH within 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.te ated Leak Rat est Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for thc last 24 hours.Excessive leakage ie indicated by a 82 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for drywell tcmperaturc, pressure, and venting operations) at 49.6 peig.Corrected to normal dryvcll operating pressure of 1.1 psigf this value is 542'SCFH.If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.Perform leakage rate testing in accordance with thc Primary Containmcnt Leakage Rate Testing Program.5<<WuSW Fs'cation Q<Q,Q>9'c BC'Sf'5 3,g,~,(BFH Unit 1 3.7/4.7-3 aMreMmr NO.228.PAGE~OF Cl~, 4.7.A.2.(Cont'd)Zf at any time it is determined that the criterion of 4.7.A.2,g is exceeded, repairs shall be initiated immediately. +~'e+~kÃi~4it'Gr 6 Aptly 4~gf'N IsrS g.g/.Iyg~(p (2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours f ollowing detection of excessive local leakage, the reactor shall be shut down and depressurixed until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.sg s.6./.3./o tK isolation valves shall be tested at a pressure of 25 psi for leakage durin ach refuels.n p2.outa e Zf t e leakage rate of 1l.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 3.7/4.7-8

3.7.C 3-Secondary containmcnt integ-rity shall be maintained in the refueling xone, except as specified in 3.7.C.4.4.If refueling zone secondary containment cannot bc maintained the following conditions shall be mct: a.Handling of spent fuel and all operations over spent fuel pools and open reactor~elis containing fuel shall be prohibited. b.The standby gas treatment system suction to thc refueling zone vill be blocked~cept for a controlled leakage area sized to assure the achieving of a vacuum of at least 1/4-inch of water and not over 3 inches of~ater in all three reactor zones.This is only appli-cablc if reactor xone integrity is required.ey.cgt e Keac~c Suig g lfd culm br<ag~g~36./.3 Mhen Primary Containment Integrity is required, all rimary containment isolati valves and all reactor coolant system instrument line floe check valves s 11 bc OPERhB except as specified in 3.7.D.2.*Locked or sea cd closed valves may be opened on an inter-mittent basis under administrative control.a.~<3,e.l,3.5 CR p.g,!,~.to SP.g,v, l.3.1 At least once r o cr-ating c clc the OPER-primary contain-ment isolation valves that are po~er operated and automatically initiated shall be tested for simulated automatic initiation Ac os.Ll 1.The primary containment isolation valves surveillance s6all bc performed as follows: SFN Unit 1 3.7/4.7-17 SlegMNr N0.y 8 g

MckEI~Qai SR 3.i.1.3.5 az,c,.l z d ia accordance vith Specification lo0ol%lg tested for closure times.f<8'M HC1(op)8 t'r.e4ug 8<T>oz p (Q r sid kCr i~)F b.In accordance with Specification 1.0.%5, all normally open povcr operated primary containment isolation valves shall bc fuactionally tested.f~~s~4 AJ4~onl5 1<4 used ALIAS 3+9 H Q.TiOWX C~~<3.4.1.'3 t d EF'CV ac l4 Bc iSo)pHo PsiW~o n Sindhi tah J i tlstvu~g+/q Wcc t signai (Deleted)At leas once pcr operatin c cl the OP ILITY of the reactor coolant system instrument line floe check valves shall be verified.QTlo PyG!2.Ia the even any primary c tai olation valve becomes inoperable, reactor operation may continue provided l~nl'how at least oae valve, ia each line having an inoperable valve, is OPERhBLE and s eithers Lq 20<<Owu<A ItC Tip P~r 4t.'o~Mhenever a primary contaiamcnt isolation valve is inoperable, thc position of at least the valv each line having an ino rable valve shall be ecor e daily a.The inoperablc valve is restored to OPERABLE status, or b.Each affected line is.isolated by use of at least one deactivated contaiameat RC7(od isolation valve secured ia the isolated position.3.If Specification 3.7.D.l and 3 7.D.2 cannot be mct, aa orderly shutdcnm shall be~<<4" initiated and the reactor shall g bc ia th COLD SHUTDOWN CONDITION+it hour's~l(rtosedPs P,t,l, Z.)3o 0 el i3.Z 3.4~l.3.'3 3as ol g,s,, ls 3,g<loSed mang~l Valve>gl;+~l~<~o<<hect Vole W4~Cl~4h~~gh<he Valge Sec~rg L3 SpÃUnit 1 He HoTSNutnowu C~on->o~<n iZ+~vs@gal.iA 3.7/4.7-18 NENOMHfT N5.f 8 9 i ACi"~C-

5ftC<4<rabO 3.t..l.3 FEB 13 19%3.7.F.4.7.F..The primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3.7.F.2.a.The results of the ia-place cold DOP and hslogenatcd hydrocarbon tests at design flovs on HEPA filters and charcoal adsorber banks shall shov g 99K DOP removal and g 99K halogcnated hydro-carbon removal vhen tested in accordance vith AHSI H510-1975. b.The results of laboratory carbon sample analysis shall shov g 85K radioactive methyl iodide removal vhca tested ia accordance vith ASTN D3803.c.System flov rate shall bc shown to be vithin g lOX of design flov vhea tested ia accordance with ASSI H510-1975'e ao The 18-inch primary contain-acnt isolation valves asso-ciated vith PURQIHC may be open during the RUN mode for a 24-hour period after enteriag the RUN aodc aad/or for a 24-hour period prior to entering the SHUTDOWH mode.e OPERABILITX o BPS Unit 1 c.co Z 6.i.~3.7/4.7-21 2.If the provisions of 3.7.F.l.a, b, and c cannot be aet, the system shall be declared iaoyerable. The provisions of Technical Spccificatioa 1.C.l do aot apply.PURCIHC aay coa-thrns using the Staadby Qas Treatment Systea l.At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorbcr banks shall be demonstrated to be less than 8.5 inches of vater at system design flow rate{g 10%).a.The tests and sample analysis of Specifica-tion 3.7.F.1 shall be performed at least once per operatiag cycle or once every 18 months, vhichever occurs first or after 720 hours of system operation and following significant painting, fire, or chemical release in any ventilation zone commmicating vith the systcao b.Cold DOP testing shall be performed after each complete or partial replacement of the HEP filter bink or after any structural mainte-nance on the system housing.c.Halogenatcd hydrocarboa testing shall be performed after each complete or partial reylacement of the charcoal adsorber bank or after any structural aaintenancc on the system housing.S<.'e 5<bagl~Hon Q Qgg<g 4'r Cy5 37 F/q<7~F ln t.h4'c<hon g, (,].3 APA 29 tg9t these primary~'~'i'~containment fsolation valves is governed by Technical Specification 3.7.D.b.Pressure control of the cont nment i norma ly perfo ed by IHQ through 2-fnch rfmary ontainm t isol tion ives vh ch rout ef luent t the St dby Gas reatmen System.The EIUNILI o f these rimary contaf ent iso tion valves i governe by Technical pecification 3.7.D.Z.A (3.7.G.1.The Containment Atmosphere Dilution (CiD)System shall be OPERABLE vith: a.Tvo independent systems capable of supplying nftrogen to the dryvell and torus'.Cycle each solenoid operated air/nitrogen valve through at least one complete cycle of full travel fn accordance vfth Specification 1.0.MK, and at least once per month verify that each manual valve fn the flov path is open.BPS th6t 1 b.A minimum supply of 2,500 gallons of lfqafd nitrogen per systems<<St'.t-I C+O~Q(C~4<4".7/4.7-22 b.Veri that the CAD System contains a minimum supply of 2,500 gallons of lfgafd nitrogen tvic er veek.ENDMae NtL Z se PAGE~OF~ INSERT PROPOSED NEW SPECIFICATION 3.6.1.4 Insert new Specification 3.6.1.4,"Drywell Air Temperature," as shown in the BFH Unit 2 Improved Standard Technical Specifications. 0 JUSTIFICATION FOR CHANGES BFN ISTS: 3.6.1.4-DRYMELL AIR TEHPERATURE TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new Specification is being added requiring drywell air temperature to be 150'F.This is required since some accident analyses assum'e this temperature at the start of an accident.Appropriate ACTIONS and Surveillance Requirements are also added.This is consistent.with the BWR Standard Technical Specifications, NUREG 1433.0 BFN-UNITS 1, 2, 8L 3 Revision 0.u 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF 5'eCe.hCa Ond b f Al 4 7 A.a Co 2's LCo 3.a.l.3 PPPli Cab~t e b.C~Primary containment integrity shall be maintained at all times vhen the reactor is critical or.vhen the reactor vater temperature is above 212 F and fue s in the reacto esse ccpt w c performing"open vessel" physics tests at power levels not to exceed 5 Kf(t).Primary containment iategrity is confirmed if the maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of thc primary containment volume pcr 24 hours at the 49.6 psig design basis accident pressure, Pa.If 52 makeup to the primary containment averaged over 24 hours (corrcctcd for pressure, temperatures aad venting operations) exceeds 542 SCFH, it must bc reduced to<542 SCFH vithin 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.e ae Primary containment nitrogen consumption shall bc monitored to determine the average daily'nitrogen consumption for thc last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary contaiamcnt free volume per 24 hours (corrcctcd for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal dryvell operating prcssure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified ia 3.7.h.2.C shall be taken.Perform leakage rate testing ia accordance vith thc Primary Containmeat Leakage Rate Tcstiag Program.5ee&~we'eeei n CeeChuys p BF<isis 3 r.i.l~0 BPS Unit 2 3.7/4.7-3 hMENStEÃf NL 2 4 3 FAGE~OF 7 +ciCAfiow s.C./.3 4.7.A.4.7.A.2.(Cont'd)(1)If at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated immediately. ~c>$446$io~p C%~5'~gfh/IS7S 3.6.I./+3.C./.2.(2)If conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.3.C./.'R The main steamlxne isolation valves shall be tested at a pressure of 25 psig for leakage during eac re ue x If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 2 3.7/4.7-8 PAGE

3.7,C, Se o d Co ta ent 3.Secondary containment integ-rity shall be maintained in the refueling zone, except as specified in 3.7.C.4.qgP VASTl AJAR'iON FOR gpwN~s Fog ZHiv ad 3.6.V./4.If refueling zone secondary containment cannot be maintained the folloving conditions shall be met: a.Handling of spent fuel and all operations over spent fuel pools and open reactor wells containing fuel shall be prohibited. b.The standby gas treatment system suction to the refueling zone vill be blocked except for a controlled leakage area sized to assure the achieving of a vacuum of at least 1/4-inch of vater and not over 3 inches of eater in all three reactor zones.This is only appli-cable if reactor zone integrity is re uired.ms+paa J r 1b'll.l, QC,(aalesara Qrteakg r(ima V e oato ma V e tai t Isol on 1.The primary containment isolation valves surveillance shall be performed as follows: Q~Shen Primary Containment Integrity is required, all rimary containment isolation valves and all reactor coolant system instrumen line flow check valves shall be OPERAB except as specified in 3.7.D.2.~.l.g~li~ob.IA LC.O 3.lo./, 3 a.At least once per o er-ti c c the OPER-ABLE primary contain-ment isolation valves that are pover operated and automatically initiated shall be tested for imulated automatic initiation act+ad.or I NENOMEgr go.2 0 g sR 3.4.1.3.5 SR~4.I.M SR3.l'm.1.3.7 go+(~+CTIONg<gag ko SR.3.C.t)3.7/4.7-17 BFH Unit 2 PAGE t OF I*Locked or sealed closed valves may be opened on an intermittent basis under administrative control. ' Vaa'yes at L5 P~P~AC7.(oe 8 Propos 4'o6l D sR 3.g,/,3p SR~C.(.3.g and in accordance with Specification 1.0.MM, tested for closure times.In accordance with Specification 1.0.MN, all normally open power operated primary containment isolation valves shall be functionally tested.PmposM hcT(oQ p~"o~c'4 No>4.2 4a ACAahP/6 Propos~(" cs g l 4ci c7cogg c.(Deleted)Sg 3.C.l.8.8 d.tFCV w+Q 4~iso/aA'm p4S>)i~o>f s:~/~~lh jfrlf~<g/l~C$~S)vo At least nce pe o erati'c c the OPERABILITY of the reactor coolant system instrument line flow check valves shall be verified.A('TlOhLS P,e+2~CoA ki Ae c.2.~A/4<<cg A ck~4.Z+C,2.enever a primary contain-ment isolation valve is inoperable, the osition of t least one other valve in each line having an ino erable valve shall be recor e ail a.The inoperable valve is restored to OPERABLE status, or~roPoL~Sls 3 Co.I 3 3.C, L3.2S.C.J 3.3 3.4/2.g Z.C.(P.g b.gcyuirQ Aa4ia~.R.I+C.[Each affected line is isolated by use of at least one deactivated containment isolation valve secured in the isolated positio mc wag w~Ip4 f,/;g"k i o~+Ice(c.vg/P/~+J~lr~/~Secor~If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be i t COLD SHUTDOWN COHDITIOH within ours.3.AC.<lou E, AMENDMEHT NO 2 0 4 5 o;,V 3.7/4.7-18 3&BFH La Unit 2 Pl 8~'+ggacVlA~~~ogo iTic (2 lip,~In the event any pr mary conta n ent o va v becomes inoperable, reactor operation may continue provided at least one valve, in each line having an inoperable valve, is OPERABLE and thin 4 hours either: LY ~7.F.o t 4.7.F.~Ss~te 0 1.Thc primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3..7.F.2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show g 99K DOP removal and g 99K halogenated hydro-carbon removal vhen tested in accordance vith AHSI H510-1975. b.The results of laboratory carbon sample analysis shall shov g 85K, radioactive methyl iodide removal vhcn tested in accordance vith ASTI D3803.c.System flov rate shall be shown to be vithin g 1OX of design flov vhcn tested in accordance vith AHSI H510-1975~2.If the provisions of 3.7.F.l.a, b, and c cannot be met, the system shall be declared inoperable. The provisions of Technical Specification 1.C.l do not apply.PURGIHG may con-tixxae using the Standby Gas Treatment System.3o ae SR s.<1.3.I f4of~The 18-inch primary contain-ment isolation valves asso-ciated vith PURGIHG may be open during the RUH mode for a 24-hour period after entering the RUN mode and/or for a 24-hour period.prior to eater the SHUTDOMN mode.e 0 TY of BPK-~-'~->"....3.7/4.7-21 Unit 2-1.At least once every 18 months, the prcssure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 8.5 inches of vater at system design flov rate (g lOX).a.The tests and sample analysis of Specifica-tion 3.7.F.1 shall be performed at least once per operating cycle or once every 18 months, vhichever occurs first.or after 720 hoars of system operation and f olloving significant painting, fire, or chemical release in any ventilation xone communicating vith thc systems b.Cold DOP testing shall be performed after each complete or partial replacement of the HEPA~filter bank or after any structural maintc nance on the system housing.c.Halogeaatcd hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.SEE&$TI Pic'&ion)f og<<""-5$F~~c~S g 7~/q Sec4io~NENMRf NO 231

APR 2 9 1991~s.a e these primary/Co 9.C.t.3 containment isolation valves is governed by Technical Specification 3.7.D.b.Pressure control of the contai nt is normally performed by VENTING through 2-ch primary containment olation valves vhich ute ffluent to the Standby G Treatment Sy em.The PERABILITY o these rimary contai ent isolatio valves governed by Technical ecification 3;7.D.3.7.G.Co ta e mos ere ut o S ste C 1.The Containment Atmosphere Dilution (CAD)System shall be OPERABLE vith: 4.7.G.Co ta e t t os ere utio S stem C D 1.S st 0 crab t a.Tvo independent systems capable of supplying nitrogen to the dryvell and torus.a.Cycle each solenoid operated air/nitrogen valve through at least one complete cycle of full travel in accordance vith Specification 1.0.MM, and at least once per month verify that each manual valve in the flov path is open.~..Unit 2 b.A minimum supply of 2,500 gallons of liquid nitrogen per system..7/4.7-22 5 E~0<S TI F'C AT t o g F'o~C9+n'~Fag.Pp/J<<~Z.g.g,i~b.Verify that the CAD System contains a minimum supply of 2,500 gallons of liquid nitrogen tvice per veek.AMENDMEMT N0.I 9 7'AG~~o~~ 0' INSERT PROPOSED NEW SPECIFICATION 3.6.1.4 Insert new Specification 3.6.1.4,"Drywell Air Temperature," as shown-in the BFN Unit 2 Improved Standard Technical Specifications. JUSTIFICATION FOR CHANGES BFN ISTS: 3.6.1.4-DRYWELL AIR TEMPERATURE TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new Specification is being added requiring drywell air temperature to be 150'F.This is required since some accident analyses assume this temperature at the start of an accident.Appropriate ACTIONS and Surveillance Requirements are also added.This is consistent.with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2, 5, 3 Revision 0

UNIT 3 CURRENT TECHNICAL SPECIFICATION 2oae+l>CA bo l e'fy b.co Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water t'emperature is above 212 F and fuel is in the reactor vessel e ing"open vessel" physics tests at power levels not to exceed 5 MW(t).Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.If H2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to<542 SCFH within 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.te rat d e ate est Primary containment nitrogen consumption shall be monitored to determine the average daily.nitrogen consumption for the last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for drywall temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified in 3.7.k.2.c shall be taken.Perform leakage rate testing in accordance with the Prima Containment Leakage Rate Testing Program.e g~hknlfio~ kR Qagr5 foA BFQ[5+5 g,4g,/BFK Unit 3 3.7/4.7-3 NENMERr go.2 Og ma~~

4.7.A.4.7.A.2.(Cont'd)h.(1)Zf at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated immediately. ~<~+~$5~~4%40~4" C(~~Ar it/-nl/st y.g,/.I~~~I+(2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.SR Z.C./3.~o main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refuelzn outa e Zf the leakage rate of 11.5 scf/hr f or any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 3 3.7/4.7-8 0 3.7.C 3.5ccondary contaiameat integ-rity shall bc maintained in the rcfueliag zone, except as specified in 3.7.C.4.4.Ig refueliag zone secondary containment cannot be maintained the folloving conditions shall bc met: cc wushCi'ccgi~ Ch~~, 8c'N isfs w.~.q,j a.Handliag of spent fuel and all operations over spent fuel pools and open reactor vclls coataining fuel shall bc prohibited. The standby gas treatment system suction to the refucliag xone vill bc blocked except for a controlled leakage area sized to assure the achieviag of a vacuum of at least 1/4-inch of vater aad not over 3 inches of vater in all three reactor zones.This is only appli-cable if reactor zone integrity is required.epee~geaH+v pcgguW+<at'c<<A pcVion F b.In accordance vith Specification 1.0.MN, all normally open pover operated primary containment isolation valves shall be'unctionally tested P cnuz A+~2.eoAcii lion A+e.PapoScd Noh W 4 Rch'on5~oP<Srd N&S 3+/+AonS In the even primary contain ent isolation valv ccomes o c, reactor operation may continue provided at least one valve, in each line having an inoperablc valve, is OPERABLE and our either: The inoperable valve is restored to OPBRABLB status, or%H n 4,+C inoperable, the position of at least one valv in each line having an ino erablc valve shall be ordcd da y.f Oops'tl SRs 3 c.(Deleted)5 Z d.At least once per operating cycl the ECe4 a Ww oPB o the fo%t<5ol atia&reactor coolant system fes:how on a instrument linc flov check valves shall be verified linc ggav.S ignis 2.whenever a primary contain-ment isolation valve is b EcQa6rclk 4@+n g,l+4al Each affected linc is isolated by use of at least one deactivated containment isolation valve secured the isolated position.3.If Specification 3.7.D.l and 3.7.D.2 cannot bc mct, an t orderly shutdovn shall bc initiated and the reactor shall G bc th COLD SHUTDOWN(COHDITIOK vi hours'i BPK 3.7/4.Unit 3<AT>Halo~Coa)Dido&LR~s 4&l~AI 7-18 L3~4.I.S.3 3'o c,l,g,g.l.3.$AMENDMENT NO.I 6 y PAGE<losed~n~)Value,gl;nd ~<a%<,~caw.i va>n~,~ohe vaNe gecur~ot Q'e cia'I~3 FEB 1 81995.7.F.t c 4.7.F.~Ssteg 1.The primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3.7.F.2.a.The results of thc in-place cold DOP and halogenated hydrocarbon tests at design flovs on HEPA filters and charcoal adsorber banks.shall shov g 99Z DOP removal and g 99Z halogenated hydro-carbon removal vhcn tested in accordance vith AHSI H510-1975. k b.The results of laboratory carbon sample analysis shall shov g 85Z radioactive methyl iodide removal vhcn tested in accordance vith AS'3803~c.System flov rate shall bc shovn to be vithin g 10Z of design flov vhcn tcstcd in accordance vith AHSI H510-1975~3o ao SC p,r,~,p,I gobe, The 18-inch primary contain-aent isolation valves asso-ciated vith PURGIHG may be open during the RUH mode for a'4-hour period after entering the RUH mode and/or for a 24-hour period prior to entering the SHUTDOWH mode.e OPERABILITX of 2.If the provisions of 3.7.F.l.a, b, and c cannot be met, thc systea shall be declared inoperable. The provisions of Technical Specification 1.C.1 do not apply.PURCIHC may con-tinue using th>>Standby Cas Treatment System.l.At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorbcr banks shall bc demonstrated to bc less than 8.5 inches of vater at system design flov rate (g 10Z).a.The tests and sample analysis of Specifica-tion 3.7.F.1 shall be performed at least once per operating cycle or once every 18 months, vhichever occurs first or after 720 hours of system operation and folloving significant painting, fire, or chemical rcleasc in azar ventilation zone communicating vith thc system.~b.Cold DOP testing shall bc performed after ea complete or partial replacement of the HEPA filter bank or after any structural mainte-nance on thc system housing.c.Halogenated bydrocarbon testing shall be performed after each complctc or partial replacement of the charcoal adsorbcr bank or after axe structura aaintcnance on the stem housing.St:e<icsAfrczgc,n Q.C+g~~~/>>1</V VFin P.kj)Section BFI Unit 3 Lco F,y.[,y AMENDMENT NO.I S 8 3.7/4.7-21

5(ec.i Ac~+on 3,4~(3 2 9 1991 3.7.F.these primary containment isolation valves is governed by Technical Specification 3.7.D.b.Pressure ontrol of thc ontainmcn is normally p rformed b VEHTIHG th ough 2-in primary con ainment is ation valv s vhich ro e efflu t to the andby Gas Tr tment Sys The OPE ILIA of thcsc pr ry containm isolation valves is vcrned by Tcchnical Specification 3.7.D, 3.7.G.4.7.G.The Containment Atmosphcrc Dilation (CAD)Systea shall be OPERABLE vith;a.Two independent systems capable of supplying nitrogen to the dryvell and torus>>a.Cycle each solenoid operated air/nitrogen valve through at least onc complete cycle of full travel in accordance vith Specification 1.0.MM, and at least once per month verify that each manual valve in thc flov path is open.BFI Unit 3 b.A miniama supply of 2,SOO gallons of liquid nitrogen pcr system>>~'~~~Auto~ro, gp cs~&It gpN ($y~7/4.7-22 b.Verify that thc CAD Systea contains a minimum supply of 2,500 gallons of liquid nitrogen tvici ck cr vc SWIDNrryy, g gs INSERT PROPOSED NEW SPECIFICATION 3.6.1.4 Insert new Specification 3.6.1.4,"Drywell Air Temperature," as shown in the BFN Unit 2 Improved Standard Technical Specifications. 0' JUSTIFICATION FOR CHANGES BFN ISTS: 3.6.1.4-DRYWELL AIR TEMPERATURE TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new Specification is being added requiring drywel1 air temperature to be 150'F.This is required since some accident analyses assume this temperature at the start of an accident.Appropriate ACTIONS and Surveillance Requirements are also added.This is consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS I, 2, 5 3 Revision.0 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.3-,.PRINRY CONTAINNENT ISOLATION VALVES ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance'ith the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.In addition, the PCIV LCO now exempts the reactor building-to-suppression chamber vacuum breakers and scram discharge volume vent and drain valves since they are governed by other LCOs.Any changes to the requirements for these valves are discussed in the new LCO Justification for Changes.A2 The current technical specification (CTS)4.7.D.l.a frequency of"once per operating cycle" has been changed to"In accordance with the Inservice Testing Program" for proposed SR 3.6.1.3.5 (stroke time tests).The CTS 4.7.D.l.d frequency of"once per operating cycle" has been changed to"18 months" for proposed SR 3.6.1.3.8.Since an operating cycle is 18 months and the current IST program requires testing every 18 months, this change is considered administrative in nature.The CTS 4.7:A.2.i frequency of"each refueling outage" has been replaced with"in accordance with the Primary Containment Leakage Rate Testing Program" for SR 3.6.1.3.10.This program requires Appendix J requirements to be met.The Appendix J requirements will always supersede the Technical Specification requirements (unless an exemption is approved)since Appendix J is the rule.Therefore, this change is purely an administrative preference in presentation. A3 This proposed Note (" Separate Condition entry is allowed for each penetration flow path")provides explicit instructions for proper application of the actions for Technical Specification compliance. In BFN-UNITS 1, 2,&3 Revision 0

JUSTIfICATION FOR CHANGES BFN ISTS 3.6.1.3-PRIMARY CONTAINMENT ISOLATION VALVES conjunction with the proposed Specification 1.3-"Completion Times," this Note provides direction consistent with the intent of the existing Actions for inoperable isolation valves.A4 The proposed ACTIONS include Notes 3 and 4.These Notes facilitate the use and understanding of the intent to consider any system affected by inoperable isolation valves, which is to have its ACTIONS also apply if it is determined to be inoperable. Note 4 clarifies that these"systems" include the primary containment. With proposed LCO 3.0.6, this intent would not necessarily apply.This clarification is consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered an administrative presentation preference. AS The current single Action for"any primary containment isolation valve" has been divided into three ACTIONS.Proposed ACTION A for one valve inoperable in a penetration that has two valves, proposed ACTION B for two valves inoperable in a penetration that has two valves, and proposed ACTION C for one valve inoperable in a penetration that has only one valve.All technical changes are discussed elsewhere in this section.As such, this change is considered an administrative presentation preference. TECHNICAL CHANGES-MORE RESTRICTIVE Ml CTS 3.7.D.3 requires an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION within 24 hours when certain conditions can not be met.Proposed Action E will require the plant be in MODE 3 in 12 hours and MODE 4 in 36 hours.The addition of this intermediate step to the COLD SHUTDOWN CONDITION is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.M2 CTS applicability for PCIV operability is when primary containment integrity is required.Per CTS 3.7.A.2.a, primary containment integrity is required at all times the reactor is critical or when the reactor water temperature is>212'F and fuel is in the reactor vessel.The proposed applicability of MODES 1, 2, and 3 is more restrictive since CTS does not require primary containment integrity when in NODE 2, not BFN-UNITS 1, 2, 8E 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.3-PRIMARY CONTAINMENT ISOLATION VALVES 0 critical and (212'F.The proposed Specification is also applicable when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, which adds a MODE 4 and 5 requirement for the RHR Shutdown Cooling isolation valves.An appropriate ACTION has been added (proposed ACTION F)for when the valves cannot be isolated (since the unit is already in MODE 4 or 5, the current actions provide no appropriate compensatory measures). ACTION F requires the licensee to initiate action to suspend operations with the potential for draining the reactor vessel immediately and to restore valve(s)to OPERABLE status immediately. If suspending an OPDRV would result in closing the RHR Shutdown Cooling valves, an alternative required action is provided to immediately initiate action to restore the valves to OPERABLE status.M3 New Surveillance Requirements have been added.SRs 3.6.1.3.1, 3.6.1.3.2 and 3.6.1.3.3 ensure PCIVs are in their proper position or state.SRs 3.6.1.3.4 and 3.6.1.3.9 ensure the traversing incore probe (TIP)squib valves will actuate if required.These SRs are additional restrictions on plant operation. This change adds acceptance criteria to the Surveillance Requirement'hich requires an Operability test of the instrument line excess flow check valves (EFCVs).The acceptance criteria added requires that the EFCVs actuate to the isolation position on a simulated instrument line break signal.The addition of acceptance criteria which did not previously exist in Technical Specifications constitutes a more restrictive change.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic".LA1 CTS 3.7.F.3.b provides no requirements, it just explains that the normal method of containment pressure control is through 2-inch PCIVs, which route effluent through the SGTS.Since the OPERABILITY of these valves is governed by proposed BFN ISTS 3.6.1.3, the specification provides no requirements and has been eliminated. Any details relating to PCIV operability have been relocated to the Bases of LCO 3.6.1.3.Placing these details in the Bases provides assurance they will be appropriately maintained since changes to these details will require a 50.59 evaluation. BFN-UNITS 1, 2,&3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.3-PRIMARY CONTAINMENT ISOLATION VALVES"Specific" Ll The phrase"actual or" in reference to the automatic isolation signal, has been added to the Surveillance Requirement for verifying that each PCIV actuates on an automatic isolation signal.Thi's allows satisfactory automatic PCIV isolations for other than Surveillance purposes to be used to fulfill the Surveillance Requirements. Operability is adequately demonstrated in either case since the PCIV cannot discriminate between"actual" or"simulated". L2 The provisions of the"*" Note of CTS 3.7.D.1 are encompassed by Note 1 to the ACTIONS, which allows penetration flow paths to be unisolated intermittently under administrative controls (except for the 18-inch purge valve penetration flow paths).However, the ISTS allowance applies to all primary containment isolation valves (except for 18-inch purge valve penetration flow paths)not just locked or sealed closed valves.The allowance is presented in proposed ACTIONS Note 1 and in SR 3.6.1.3.2, Note 2.Opening of primary containment penetrations on an intermittent basis is required for performing surveillances, repairs, routine evolutions, etc.CTS 3.7.D.2.b allows isolating the primary containment penetrations with at least one deactivated valve secured in the isolated position when one PCIV is inoperable. The proposed ACTIONS A and C of LCO 3.6.1.3 allow the use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve (for Condition A only)with flow through the valve secured.The Action utilizing a deactivated automatic or manual valve is appropriate on the basis that these isolations present a boundary which is not affected by a single failure.The ability to utilize the valves downstream of the outboard PCIVs is an acceptable isolation since it meets the acceptance criteria of not being affected by a single active failure.L4 CTS 3.7.D.2 allows reactor operation to continue when any PCIV becomes inoperable provided that at least one valve in each line having an inoperable valve is operable and within 4 hours the affected line is isolated or the inoperable valve is restored to OPERABLE status.Based on the wording, this only applies to lines with two isolation valves.This is equivalent to proposed ACTION A, however, the proposed ACTION allows additional time to isolate the main steam lines.A Completion Time of 8 hour s for the MSLs allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the MSLs and a potential for plant shutdown.For BFN-UNITS 1, 2, 8L 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.3-PRINRY CONTAINMENT ISOLATION VALVES penetration flow paths with only one PCIV, proposed ACTION C allows 4 hours to restore an inoperable valve to OPERABLE status and 12 hours to restore EFCVs in reactor instrumentation line penetrations. The four hour Completion Time is reasonable considering the relative stability of the closed system to act as a penetration boundary and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3.The Completion Time of 12 hours is reasonable considering the instrument and the small pipe diameter of the affect penetr ations.During the allowed time, a limiting event would still be assumed to be within the bounds of the safety analysis, assuming no single active failure.Allowing this extended time to potentially avoid a plant transient caused by the immediate forced shutdown is reasonable based on the probability of an event and does not represent a significant decrease in safety.L5 In the event both valves in a penetration are inoperable, the existing Specification, which requires maintaining one isolation valve OPERABLE, would not be met and an immediate shutdown is required.The proposed ACTION (ACTION B)provides 1 hour prior to commencing a required shutdown.This proposed 1 hour period is consistent with the proposed BWR Standard Technical Specification time allowed for conditions when the primary containment is inoperable. The proposed change will provide consistency in actions for these various containment degradations. L6 The frequency of the periodic verification required when a penetration has been isolated to comply with current Specification 3.7.0.2 has been changed from daily to monthly.These valves are strictly controlled and are operated in accordance with plant procedures. Daily verification that these valves are still isolated places an undue burden on plant operations and provides little if any gain in safety, since these valves are rarely found in the unisolated condition, once closed.Note that CTS 4.7.D.2 requires the position of one other valve in the line be"recorded" daily versus the ISTS wording of"verified." ISTS also allows an inoperable valve to be used for isolating the penetration. L7 The Note to SR 3.6.1.3.1 allows the SR to not be met (i.e., do not have to verify closed)when the valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry and for Surveillances that require the valves to be open.For these reasons, it is deemed acceptable to open the valves for short periods of time.CTS 3.7.F.3.a, which allows the 18-inch primary containment isolation valves associated with PURGING to be open during the RUN mode during a 24-hour period after entering the RUN mode and/or BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.3-PRIMARY CONTAINMENT ISOLATION VALVES for a 24-hour period prior to entering the SHUTDOWN mode, is encompassed by the provisions of the Note.The additional exemptions allowed by the Note are acceptable since the 18-inch purge valves continue to be capable of closing in the environment following a LOCA.L8 The time allowed to shutdown the plant when the required actions are not'et has been changed from"in the COLD SHUTDOWN CONDITION within 24 hours" to in MODE 3 (Hot Shutdown)in 12 hours and MODE 4 (Cold Shutdown)within 36 hours.The proposed 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.The additional 12 hours allowed to reach Mode 4 is offset by the safety benefit of being subcritical (MODE 3)in a shorter required time.L9'his change adds proposed ACTION D which relaxes the allowed outage time from 4 hours to 8 hours to isolate the affected penetration if one main steam isolation valve (HSIV)in one or more penetrations is inoperable (due to leakage or other reason).This will allow a longer period of time to restore the HSIVs to OPERABLE status in order to prevent the potential for a plant shutdown by isolating the main steam line(s).During the additional time allowed, a limiting event would still be assumed to be within the bounds of the safety analysis, assuming no single active failure.Allowing this extended time to potentially avoid a plant transient caused by a plant shutdown is reasonable and does not represent a significant decrease in safety.t BFN-UNITS 1, 2, L 3 Revision 0 4 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 4 c., Cq on 76~t~~JUL 17 5%Rg(lie b'lip 8c)oem'SR 3.~.t.C'crloitj 5 Aec,~t)t'f~~)far toH$B u+<ce t gs Itpcc fi d i.7.A 3.g bello vo pressure suppress on ber-reactor building vacuum breakers shall bc OPERABLE at all times vhcn PRIMARY COHTAIHMEHT IHTEQRITY is required.The setpoint of the differential pressure instzlwcntatfon vhich actuates the pressure suppression chamber-reactar hi~tLag vacuum breakers shall be pcr Table 3.7.A.trios'om and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to bc inoperable for any reason, reactor operation is permissible oaly duriag the succeeding sevca days, provided that the repair procedure does not violate PRIMARY COHTAIHMEHT IHTE DURITY.3.a.l,5.3 assoc ate instrumentation including sctpoint shall be functionally tcstcd for proper operation per Table 4 7.h Sfz.~.t.S. ~b.k s determination that the force required to open each vacuum breaker (check valve)does not cxcced 0.5 psid vill be made each refueling ou'tage e fop et SR a.The pressure suppression chamber-reactor building vacuum breakers shall be exercised in'ccordance vith Specification 1.0.MM an e 4e 4.0 ae b When primLry contaiament is required, all dryvell-suppression chamber vacmm brgakera shall be OPERABIZ and positioned in the fully closed position (except daring testing)except as specified tn 3,7.L 4.b and 3.74A.c., belor.One dryve11-suppression chamber vacuum breaker may be nonfully closed so long as it is determined to be not more thea 3 open as indicated by the position lights.a.Each dryvell-suppression chamber vacuum breaker ahall be tcstcd in accordance vith Specification 1.0.MM.b.When it is determined that ceo vacmm breakers are inoperable for opening at a time vhen OPERABILITY is required, all other vacmm breaker valves shall be exercised immediately and every 15 days thcrcafter until the inoperable valve haa tete rmal service BFR Unit 1$.7/4.7-10 AMENbMEÃf N, 2 2 2 PAGE PP~

TABLE 3.7.A 1NSTRNKNTATIOH FOR CONAlliiENT SYSTENS Minima No.Operable Per Irhdeta 1nstnaent annel-Pressure s ppresslon chamber-r actor building vacua b eakers (Pdl 20, 21).5 psld Actuat the pressure suppre sion chaiber-react r building vacu breakers.Footnote: gapa)r n 2I hours.lf the fun ion ls not OPERABLE ln 4 hours, declare th systea or coeponent lnoperabl~.

TABLE 4.7.A CONTAI T SYSTEN INSTRlNENTATI SURVEILLANCE REggIRENENTS Ins t~nt hannel-Pressure ppression chaebe~actor building vacw~o eakers (PdIS 20, 2i)Once/aonth~l) Once/1d oonthsI)n r No Footnotes: g into th electronic trip rcuitry in place of sensor gnal to verify OPKRASI ITY of the trip and alara functions. (2)-Calibr tion consists of the a justwent of the pri ry sensor and as ciated coeponent so that thev correspond within acceptable range and ccurac to%nolan val es of the parsee r vhich the chan el monitors, fncludin so g at its output r ay changes state at or core conservatively than ad]ustjaent of the electroni trip circuitry, the analog equivalent of th level setting.(I)~)-Function teat consists of the n ection of a siaul ed si nal the

UNIT 2 CURRENT TECHNICAI SPECIFICATION MARKUP ~k.L%'q BODES l~g$g 3.6.l.5 3 Aa t as~cifieL in elo tvo pressure suppression chamber-reactor building vacuum breakers shall be OPERABLE at all times vhen PRIMARY COHTAIHME2C IHTEGRITY is required.Thc setpoint of the differential pressure hzstrumcntation vhich-actuates thc prcssure suppression chamber-reactor building vacuum breakers shall bc Table 3'l.L.e s 5'F 4 c.l,S:3 associate nstrumcntation including sctpoint shall be functionally tested for proper operation per Table 4 V.A 5R'3.a.The pressure suppression chamber-reactor building vacuum breakers shall be exercised in'accordance vith Specification 1.0.MM and t e Pre Sae0 Ns4C~At-'TIO b.From and after the date Li P,c.riadS LI p~p~Ac.vroWS 8 5+8 that one of thc prcssure suppression chamber-reactor building vacuum breakers ia made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days, provided that thc repair procedure does not violate PRIMARY COHTAIHMEHT IHTEQRITY. 3.S.l.5>v s examXaacio~ determination that the force required to open each vacuum breaker (check valve)does not exceed 0.5 psid vill bc made each refueling outage.'P~p~SL 3.c.J.<I 4, 4, ao b.Shen primary containment is required, all dryvcll-auppression chamber vacuum breakers shall be OPERABLE and positioned in thc fully closed position (except during testing)except aa specified in 3.7.A.4.b and 3.7.L.4.c. ~belce.Onc dryvell-suppression chamber vacuum breaker may bc nonfully closed so long aa it ia determined to be not more thar 3 open as indicated py the position lights.a.Bach dryvc11-suppression chamber vacuum breaker shall be tested in accordance vith Specification 1.O.MM.b.When it ia detenained that tvo vacmm breakers are inoperable for opening at a time vhen OPERABILITY ia required, all other vacuum breaker valves shall bc cxerciscd immediately and every 15 days thereafter until the, inoperable valve has been returned to normal service.BFS Unit 2 3.7/4.7-10 Sce ZusAC;w4gg~ t."4~~~+4~PF<ls7S g.g.l.g hMENDMBfT HO.2 3 7.a.-.E~OF~ I TABLE 3.7.A INSTRUNENTATION FOR CONTAIISENT SYSTENS Hlnlaaa No.Operabl~Per rk Instrument Ch nel-Pressure su ress)on chamber-re tor build(ng vacuun b akers (PdIS 20, 21)0.5 paid Actuates the pressure~suppression chamber-reactor building vacua breakers.Footnote;(l)-Raper In 24 hours.If the function$s not OPERABLE)n 24 hours, declare the systen or coeponent (noperable.

TABLE 4.1.A CONTAIfRiENT SYSTEH INSTRlNENTAT ION SURVEILLANCE REgUIRENENTS n r n h k Instrueent Channel-Pressure suppression chaeber-reactor building vacuum breakers (PdlS<4-20, 21)Onc/month Once/18 aonths None Footnotes: -+I I 1J iver-Functional est consists of the injection of a sieulat d signal into the electronic tr circuitry in place of the (1)sensor sig al to verify OPERABILITY of the trip and alarm functions. ()-Calibrati consists of the ad)ustment of the prleary sensor and associated cenponents so that the~correspond (2)within a eptabl~range and accuracy to known values of the paraeeter which the channel aonltors, ncluding adJusteent of the electronic trip circuitry. so that its output relay chaqges state at or sore conservatively than the analog equivalent of the level setting. UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF~

sfec4iM'.6 l 8 JUL 17 Q5 VPlimb;);RCho~5 t A c-pu 8 Or+E.a.cept spec~ed~oAe3eb clov vo Prcssure suppression chamber-reactor building vacuum breakers shall be OPERABLE at all times vhcn PRIMhRY COHTAIHMEHT IHTEQRITY ia required.The aetpoiat of tha differential prcssure fastnmcntatf on vhf ch actuates the pressure suppression chamber-reactor building vacuum breakers shall be l per Table 3.7.A.Ao pn b.rom and after thc date that onc of the pressure suppression chamber-reactor building vacuum breakers is made or found to be iaopcrable for any reason, reactor operation fs permissible only during the succecdiag seven days, provided that the repair procedure does not violate PRIllkRT COHTAINHEHT IHTECRITY a.The pressure suppression chamber-reactor building vacuum breakers shall be mcrciscd in accordance vith Specffication 1.0.MM d the<<gt I et eeoc ated ineteueentation including setpoint shall be fmctfonaXXy tested for proper operation per Table 4.7.A.5g g.e.t..b.A ion an dctcrmination that thc force required to open each vacuum breaker (check valve)docs not exceed 0.5 psid vill be made each refueling Qu'tagce~~Posgg 8R 3 4 L 5'I 4.ess 0 BPH Unit 3 ae b.When priory coatafnmeat is required, all dryvell-suppressfoa chalbcr va~breakers shaU.bc OHSABLS sad positioned fa the fully closed posftioa (except during testing)czccpt as specified in 3.7.A.4,b aad 3.7.A.4.c belov.One dryvcll-suppression chamber vaema breaker aay be nonfully closed so long as it fs determined to be not aorc than 3 open as indicated by the posftfon lights.~~HA.ca.hb~ f~4~,~SPe lST5 3,g,t,g a.Each dryvcll-suppression chamber vacuum breaker ahall be tested in accordance vith Speci ficatioa 1.0.MM.b.When it fs detezafncd that tvo vacuum brcakcrs are iaoperable for opening at a time vhen OPERABILITY is required, all other vacuum breaker valves shall be ezercised hamediately and every 15 days thereafter until the inoperable valve has been returned to 1 NPnMENT NO.19 6 TABLE 3.7.A INSTRNNTATION CONAl~Sy5Tg6 iiiniaaa No.Operable Per XrhJiuha Instrument Channel-Pressure suppression chiober<<reactor bui Ini vacua breakers (Pdl~20, 21)0.5 paid Actuates e pressur suppress on chamber-reactor uildini vacu reakers.Footnote: (1)-Repair in 21 hours.Lf the function is not OPEINBLE In 24 hours, d clare the system or component inoperabl~. TASLF.i.l.CONT igiENT SYSTEM INSTRSKNATI SURVEILLANCE REgUIRENENTS Instant Channel-Pressure suppression chaeber-reactor building vacua brea'kers (PdIS-64-20, 21)Once/senth(I~ Once/18 s>>nth n n h Footnote:: ~-Funct onal test consists f the in)ection of sisulated signal int the electronic trip circuitry place of the (11 sans si9nal to verify ERASILITY of the t p and alarm function.(~-Cal bration consists o the ad)uits>>nt of he primary sensor an associated components so that c~correspond (21 ui hin acceptable ran and accuracy to own values of the pa aa>>ter erich the channel aonito, ncluding a justa>>nt of the el tronic trip citcu ry.so that its outp t relay changes state at or aor conservatively than~analog equivale of the level set ng.tQ C> JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.5 REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS ADMINISTRATIVE CHANGES Al A2 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.Existing LCO 3.7.A.3 is being replaced by proposed LCO 3.6.1.6.The proposed LCO will contain a Note stating that: "Separate Condition entry is allowed for each line." This note clarifies that the Conditions and Required Actions that follow may be applied to each of the two reactor building-to-suppression chamber vent paths without regard to vent path status.Each vent path contains two vacuum breakers in series.This note provides directions consistent with the intent of the Required Actions.This change is consistent with NUREG-1433. TECHNICAL CHANGES-MORE RESTRICTIVE CTS 3.7.A.2.a requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.6 applicability is NODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in MODE 2, not critical and<212'F.M2 A new Surveillance Requirement has been added to verify each vacuum breaker is closed (except when they are open for performance of Surveillances) every 14 days.This is consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2, 8, 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.5 REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUN BREAKERS TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Details of visual inspections of valves have been relocated to plant procedures. This type of inspection is more appropriately controlled by plant procedures. The valves are still required by Technical Specifications to be cycled and their setpoint verified to ensure operability. Any changes to procedures will be controlled by the licensee controlled programs."Specific" Ll Existing LCO 3.7.A.3.b identifies the currently required actions if one reactor building-to-suppression chamber vacuum breaker is inoperable. If more than one vacuum breaker is inoperable, the existing specification assumes either containment integrity is lost or the ability to relieve negative pressure in the containment is lost.Therefore, LCO 3.7.A.3.b. defaults to 1.0.C.1 which requires that the reactor be placed in Hot Standby within 6 hours and Cold Shutdown within the following 30 hours.Proposed LCO 3.6.1.6 recognizes that there are two vacuum breakers in series in each of two vent paths between the reactor building and suppression chamber.As a result, if one vacuum breaker in each vent path is not closed (Condition A), containment integrity and venting capability are still maintained and 7 days is provided to restore the redundancy for containment integrity in each vent line.Likewise, if two vacuum breaker valves in one vent line are inoperable but closed (Condition C), containment integrity and venting capability are still maintained and 7 days is provided to restore the redundant vent path.Therefore, proposed Specification 3.6.1.6 makes the distinction between loss of redundancy and loss of function.The existing specification fails to make this distinction between loss of function and loss of redundancy and, therefore, is unnecessarily conservative. In addition, loss of function (loss of containment integrity (Condition B)or loss of venting capability (Condition D))will require initiating action within 1 hour instead of immediately. Also, CTS 3.7.A.3 does not have a specific shutdown requirement, therefore, CTS 1.0.C.1 applies.CTS 1.0.C.1 requires the unit be placed in Hot Standby within 6 hours and Cold Shutdown within the following 30 hours.Proposed ACTION E requires the Unit to be placed in Hot Shutdown with 12 hours and Cold Shutdown within 36 hours.Proposed ACTION E is considered less restrictive since additional time is allowed prior to requiring the plant to be in a lesser Node (i.e., Proposed Action E requirement to be in Hot Shutdown in 12 hours versus the CTS requirement to be in Hot Standby in 6 hours).This change is consistent with NUREG-1433. BFN-UNITS 1, 2, L 3 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.5 REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS L2 The vacuum breaker actuation instrumentation Surveillances are proposed to be deleted from Technical Specifications. The requirement of SR 3.6.1.5.3 to ensure the vacuum breakers are full open at 0.5 psid is sufficient. Vacuum breaker actuation instrumentation is required to be OPERABLE to satisfy the setpoint verification Surveillance Requirement (SR 3.6.1.5.3) for the vacuum breakers.If the vacuum breaker actuation instrumentation is inoperable, then the Surveillance Requirement cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the ACTIONS of Specification 3.6.1.5.As a result, the requirements for the vacuum breaker actuation instrumentation are adequately addressed by the requirements of Specification 3.6.1.5 and SR 3.6.1.5.3 and are proposed to be deleted from Technical Specifications. BFN-UNITS 1, 2, 8L 3 Revision 0 PAGE~OP~

CURRENT TECHNICAL SPECIFICATION MARKUP

3.7.A 4.7.A 3~3~aes ao Except as specified in 3.7.A.3.b bclov, tvo pressure suppression chamber-reactor building vacuum breakers shall bc OPERABLE at all times vhen PRIMARY COHTAIHMEHT IHTEGRITY is required.Thc setpoint of the differential pressure instDKcntatioa'vhich actuates'he prcssure suppression chamber-reactor ~Bag vacuum breakers shall be pcr Table 3.7.A.a.Thc prcssure suppression chamber-reactor building vacuum breakers shall be exercised in accordance vith Spccificatioh 1.0.MM, and the associated instrumentation including setpoint shall be functionally tested for proper operation per Table 4.7.k.b.Prom and after the date that one of thc prcssure suppression chamber-reactor building vacuum brcakcrs is made or found to bc inoperablc for any reason, reactor operation is permissible only during the succeeding seven days, provided that the repair procedure does not violate PRIMARY COHTAIHMEHT IHTEGRITY. b.A visual examination and dctcrmination that the force required to open each vacuum breaker (check valve)docs not exceed 0.5 psid vill be made each refueling outagee~)>C4pif>NoQP5 I,>+3.When pr containment is required all dryve suppression chamber vacma breakers shall be OPERABLE sad positioned in the fully closed positioa~ocg4-~)c n.7.4...c b er.LI:o 3,G, I.b b.When it is determined tha bre rs ar tvo vac perab for o ng t a t vhea is re ired, othe vac bre r valv shall be merc ed imm iately and every days ereaft uati thc inop rable lve ha been returaed rmal service.~h pal Q~'>>>g&Cia>hg>>>IgcI, b.One dryvcll suppression chamber vacuum breaker may bc nonfully closed so long as it is determined to be not more thm 3 open as indicated by the position lights.f JOt'~+ACTIoQ A HQ 4o jg BPH LI Uait 1 4 2, 8.7/4.7-10 hMENbMEÃf 50.2 2 2 pAGE~OF~+SR a.Each dryvell-suppression chamber vacuum breaker shall be tcstcd in accordance vith Specification 1.0.ÃM.PAc>I>c>s+ SC, 3 6>!.6>>I

c.zo 4ryvell-suyyrcssioa chamber vacuum breakers may be determined.to be yerable for op~0 gR 3.4.I ob 3 shaL)be inspected for prayer operation of the valve and 1&" wi-""cs ance v1 Syecat'n L.Q..!.~c V~s wQ~~QvC~~gp g+~Fo isvS S.I,<.l QC7(onl C-S pec'cat'oas 3.7.4.4.a, 3.7.4.4.b, or 3.7.L.4.c. cannot be met, the unit shall be placed ia a , COLS SHUTDOWN CQHDITIOH ia an orderly manner i%thin hours.In a HoT5guTDo~A 3I L4 Andi H on Jn INA<S 5~4.k Leak test of the dr.~cl to suppression chamber st~cure shaLL be conducted each operating cycle~acceptable leak rate is 0.09 Lb/sec of yrimazy coatainmeat atmosphere vith 1 psi diffe cntiaL.a.Containmeat atmosphere shall be'educed to less than 4%oxygen vith aitrogea gas during reactor~pover operation vith reactor coolaat pressure abore 100 psig, except as syecified in 3.l.i.5.b. a.The primary coataiamcat oxygea concentration shall be measured aad recorded daily.The oxygea measurement shall be ad]usted to account tor the uncertainty of the method used by adding a predetermined error funct'oa b.Within the 24-hour yeriod subsequent to ylaciag the reac or ia the RUH NDE folloviag a shut-dova, the coataiameat atmosphere oxygen coacentrat'oa shall be reduced to Less than 4" by volume aad maintained ia this condit'oa. Deiaertiag may commence 24 hours prior to a shutdovn.b.The methods used to measure the yrimary contaiameat oxygen concentration shaLL be calibrated once every refueling cycle.c.If plant control air is being used to suyyly the pneumatic coatrol system~ide primary containmcat, the reactor shall not be started, or if at, pover, the reactor shall be brought to a COLD SHUTDOM5 CQHDI OH vithia 24 hours.c.The coatrol air suyply valve for the pneumatic control system'inside thc prmar/containment shall bc vcr'f'cd closed prior to:eac or star='aad monthLy thercait.".4.If Specification 3.7.h.S.a and 3.7.A.5.b cannot be met, aa orderly shutdova shall be initiated and the reactor shall be ia a COLD SHGTDOMH COHDITIOH vithia 24 hours.~'"~urn A r Oramong RR 8FAl (5T5 AMENDMENT NO y~9 BFH Unit 1.7/4.7-11 3 r'g

UNIT 2 CURRENT TECHNICAL SP ECIF ICATION MARKUP 0 0 s.v.a 4.7.h 3~3~Su b.Pht+f ltC4~e t<pcibEs ao/~L f t c,o z.c,(.6 M<n pW4g i'4c'~;~~Curbsh b.PoQ 2.sR z.c,.l,b.l From and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to be inoperable. for any rcuon)reactor operation is permissible only during thc succeeding seven days, provided that the repair procedure does not violate PRINhRY COHThIHMEHT IHTEQRITY. When primary conta ent is e ed all dryvell-suppression chamber vacuum breakers shall be OPERABIS and positioned in the fully closed position 4emeege)exc tas.7.h.i.P l aps 371.4~OILc dryvcll suppression chamber vacuum breaker msy bc nonfully closed so long as it is determ~to be not more tban 3 open as indicated by the posit'ion lights.Except as specified in 3'ohe3eb beloved tvo pressure suppression chamber-reactor building vacuum breakers shall bc OPERhBLB at all times vhen PRIORY COHThIRNEHT IHTEGRITY is required.The setpoint of the differential pressure imtnacntation Mch-actuates.the pressure suppression chamber-reactor building vacuum breakers shall be yes Table 3'7.k.a.The pressure suppression chamber-reactor building vacuum breakers shall be exercised in accordance vith Specification 1.0.MM, and the associated instrumentation including setpoint shall be functionally tested for proper opcratioa par Table-4 ZA b.h visual examination and determination that the force required to open each vacuum breaker (check valve)does not exceed 0.5 psid vill be made each refueling outagce Sqe a~a4'i/i'(aA'oe For C4o~pSX g~BpN (sos a.C.(,g Al sp s.c./.t"~chamber vacuum breaker shall bc tested in accordance Pl vith Specification 1.0.MM.Pr~sg, p, g.(.g (en t s e e ed that tvo vacma b a are inoperable for at, t shen 0 is r rsd, all other cd br valves shall b~exercis immediately and~every 15 s thereafter until the inoperab valve has been returned to normal service.BPS~~op~4erroN A Unit 2 LZ~"0~pc~<<~8 3.7/4.7<<10 Nettegpgg. 2p q PAGE W GF S.li ea lichen 3, g.I.6 1S88 tt=l~I Co3(.lk AfVIoH 5.0 c.Tvo dryvell-suppression chamber vacuum breakers may be determined to be inoperable for opening.g.s s<~~~co~a 4 a d.If Speci f ications 3.7.A.4.a,.b, or.c cannot be met, the unit shall be pl'aced i Cold Shutdovn condition in an orderly manner vithin ours~96 4~C~5.0~I, 3 Each vacuum breakc valve shall be inspected'or proper operation of the valve and limit sv'tches I in accordance vxt SgpawsTi<iCATi4~ cification 1.0.:R&R-CII"+~~II f N sS 7 S 3.4,.I I A e test of the d./well to suppression chamber structure shall be conducted during each operating cycle.Acceptable leak race gs 0.09 lb/sec of primary containment atmosphere with 1 psi differential. Co tao Containmeat atmosphere shall be reduced to less than 4X oxygen vith nitrogen gas during reactor power operation vith reactor.coolant prcssure above 100/psig, except as specified in 3.7.A.5:b. a.The primary containment oxygen concentration shal'e measured aad recorded daily.The oxygea measurement shall be adjusted to accouat'for thc uncertainty of the method used by adding a predetermined error function.b.Mithin thc 24-hour period subsequent to placing the reactor in the BUH mode folloving a shut-dovn, the containment atmosphere oxygen concentration shall bc reduced to less than 4X by volume and maintained in this coaditioa. Deinerting may commence 24 hours prior to a shutdova.b.The methods used to measure the primary containment oxygea concentration shall bc calibrated once every refueling cycle.c.If plant control air is being used to supply the pneumatic coatrol system inside primary coatainment, the reactor shall aot be started, or if at power, the reactor shall bc:brought to a Cold Shutdown condition vithin 24 hours.c.The control air supply valve for the pneumatic control system inside the primary containment shall be verified closed prior to reactor startup and monthly thereafter. d.If Specificatioa 3.7.A.5.a aad 3.7.A.5.b cannot bc met, an orderly shutdovn shall be initiated and the reactor shall be in a Cold Shutdown condition vithin 24 hours.SFE'us7 IFICA7innJ F'og QHA>GES'oR g~nl i~3.4.B,J BiH Unit 2 3.7/4.7-11 Mi'40M'.ir. i 5-PPQL op~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF 3.7.h 4.7.h 3~ambe-aa Except as spccificd in 3.7;h.3.b bclov, tvo prcssure suyprcssion chamber-reactor building vacuum breakers shall be OPERABLZ at all times vhcn PRIMARY COHTAIHMEHT IHTEt RITY is required.The aetpoint of tha difforcaaf.al prcssure inatrL<<lcntation vhich actuates the prcssure suppression chamber-reactor building vacuu<<breakers shall be per Table 3.7.h.1zuU<era a.The prcssure suppression chamber-reactor building vacuum breakers shall be exercised in accordance vith Specification 1.0.MN, and the associated instrumentation including sctpoint shall be fictionally ecstcd for prop operation per Table 4.7.h..From and after the date that one of the pressure suppression chamber-reactor building, vacuum breakers ia<<adc or found to be inoperablc for any reason, reactor oyeration ia permissible only during the succeeding seven days, yrovided that thc repair procedure does not violate PRIMARY C02ITAIHMENX'ITECRITX, b.h viaual examination and determination that the force required to open each vacuum breaker (check valve)does not exceed 0.5 psid vill be made each refueling outagee 5'~<5uSACi cab'on Qr ChanyrZ+a di<l S Ts 3.c.!-5 4.>aabalkj a OOeS lp.<LM i@hen~get~~~ir tn~/}b.BFK Unit 2%hen pri<<ary contain<<ent ~required auypreaaion cha<<ber vacmm braakcra shall be OPERJBIS and positioned in the fully!\Opt as pec.7.l.4~1 and.7.A..bclov.3'/4.7-10 One dryvell-suppression chamber vacmm breaker<<ay be nonfully closed so long aa it ia determined to be not<<orc than 3 oycn aa indicated by the position lights.flo OQCA ftg70n, ro iong ch'on se 3...c-.a.Ea dryvcll-suppression chamber vacuum breaker shall be tested in accordance vith Specification 1.0.MN.tl C I.C.b.Shen t a deter<<ined that two va breakc are inopcrabl for open ng at a ti<<e vhen ILI is squired~other v uum b esker valv shall b ex ciaed imm ately eve 15 days th reafter til the in erablc va ve has ecn return to normal ervice NB:AMENT No.~9 6 p,a GF:~QF~ ~p>~~c.Tvo drywell-suppression chamber vacuum breakers may.be determined to be inoperable for opening.5 3'.t'o.I.4~C~3 Once each operating cycle, each vacuum breaker valve shall be inspected for proper operation of the valve and limit svitches n accordance v t Specification 1.0.MM..d.If Speci f i cations 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c, cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner vithin rs~3'CA/f 584TSecP o~'hen/n wars a.Containment atmosp ere shall be reduced to less than 4X oxygen vith nitrogen gas during reactor pover operation vith reactor coolant prcssure above 100/psig, except, as specified in 3.7.A.5.b. d.a.h leak test of the drywell to suppression chamber structure shall be conducted during each operating cycle.Acceptable leak rate is 0.09 lb/sec of primary containment atmosphere vith 1 si differential ye V~S.hQ'aCHO ~~~f r so~LSqs R.r The primary containmen oxygen concentration sh 1 be measured and recorded daily.The oxygen measurement shall be ad)usted to account for the uncertainty of the method used by adding a predetermined error function b..Wi,thin the 24-hour period subsequent to placing the reactor in the RUH mode folloving a shut-dovn, the containment atmosphere oxygen concentration shall be reduced to less than 4X by volume and maintained in this condition. Deinerting may commence 24 hours prior to a shutdown.c.If plant control air is being used to supply the pneumatic control system inside primary containment, the reactor shall not be started, or if at pover, the reactor shall be brought to a Cold Shutdovn condition within 24 hours.b.C~The methods used to measure the primary containment oxygen concentration shall be calibrated once every'efueling. cycle.The control air supply valve for the pneumatic control system inside the primary containment shall be verified closed prior to reactor startup and monthly thereafter. BFH Unit 3.7/4.7-11 d.If the specifications of 3.7.A.5.a through 3.7.A.5.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition vithin 24 hours.AMEN87tlE~IT. AO.1 8 0 I3USTIFICATION FOR CHANGES BFN ISTS 3.6.1.6 SUPPRESSION-CHAMBER-TO-DRYWELL VACUUM BREAKERS ADMINISTRATIVE CHANGES Al A2 A3 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is'lready approved, adding more detail does not result in a technical change.The words"except when performing their intended function" have been added to preclude requiring the LCO to be met when the valves cycle automatically. Since their intent is to open when a sufficient differential pressure exists, this change is considered administrative only.CTS 4.7.A.4.c is performed in accordance with the Inservice Testing Program on a frequency of every operating cycle.Proposed SR 3.6.1.6.3 is to be performed every 18 months.Since an operating cycle at BFN is approximately 18 months, this.change is considered administrative. TECHNICAL CHANGES-MORE RESTRICTIVE Ml CTS 3.7.A.2.a requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.6 applicability is MODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in, MODE 2, not critical and<212'F.M2 A new Surveillance Requirement (proposed SR 3.6.1.6.1)has been added to verify the vacuum breakers are closed once every 14 days.This new SR ensures the"closed" requirement of the LCO statement is being met.This is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 PAG~( JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.6 SUPPRESSION-CSNBER-TO-DRYWELL VACUUN BREAKERS M3 CTS 3.7.A.4.d requires an orderly shutdown be initiated and the reactor to be in the Cold Shutdown Condition within 24 hours when certain conditions can not be met.Proposed Action C will require the plant be in MODE 3 (Hot Shutdown Condition) in 12 hours and NODE 4 (Cold Shutdown Condition) in 36 hours.The addition of this intermediate step to the Cold Shutdown Condition is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.TECHNICAL CHANGES-LESS RESTRICTIVE"Specific" Ll L2 L3 Proposed ACTION A allows 72 hours to restore an inoperable vacuum breaker to OPERABLE status, with one of the required vacuum breakers inoperable for opening.This is allowed since the remaining nine OPERABLE breakers are capable of providing the vacuum relief function.The 72 hours is considered acceptable due to the low probability of an event in conjunction with an additional failure in which the remaining vacuum breaker capability would not be adequate.Proposed Action B allows a short time to close an open vacuum breaker since there is low probability of an event that would pressurize primary containment. An open vacuum breaker allows communication between the drywell and suppression chamber airspace and, as a result, there is the potential for suppression chamber overpr essurization due to this bypass leakage if a LOCA were to occur.If vacuum breaker position indication is not reliable, an alternate method of verifying that the vacuum breakers are closed is to verify that a differential pressure of 0.5 psid between the suppression chamber and drywell is maintained for I hour without makeup.The required 2 hour Completion Time is considered adequate to perform this test.Existing Specification 4.7.A.4.b requires that"When it is determined that a vacuum breaker is inoperable for opening at a time when operability is required, all other vacuum breakers shall be exercised immediately and every 15 days thereafter until the inoperable vacuum breaker has been returned to normal service." This requirement is not included in NUREG-1433 and will be deleted.This change eliminates the requirement to demonstrate the OPERABILITY of the redundant vacuum breakers whenever a vacuum breaker is declared inoperable. This change acknowledges that the inoperability of a vacuum breaker is not automatically indicative of a similar condition in the redundant vacuum breakers unless a generic failure is suspected and that the periodic BFN-UNITS I, 2, 5 3 2 Revision 0=-'".F 2 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.6 SUPPRESSION-CHAMBER-TO-DRYWELL VACUUM BREAKERS frequencies specified to demonstrate OPERABILITY have been shown to be adequate to ensure equipment OPERABILITY. Therefore, this change allows credit to be taken for normal periodic surveillance as a demonstration " of OPERABILITY and availability of the remaining components and reduces unnecessary challenges and wear to redundant components.. This change is consistent with NUREG-1433. 'L4 CTS 3.7.A.4.d requires the unit to be placed in a Cold Shutdown condition in an orderly manner within 24 hours.Proposed ACTION C is less restrictive since it requires the unit to be placed in MODE 3 (Hot Shutdown condition) in 12 hours and in MODE 4 (Cold Shutdown condition) in 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.BFN-UNITS 1, 2, 8L 3 Revision 0 PAGE~OF~ UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP Cp 5Pec.,loot.o Z (.Z i~7 4.7 ab Appli s to the o rating st us of the primary an secondary contai ent system plies to th primary and s ondary conta nment in grity.~Q(~~v To assure t e integrity f the primary and econdary containment stems.To veri the integrity of the primary d secondary containment. l.At any time that the irradiated fuel is in thc AppVcab>)<ky reactor vessel, and the nuclear system is pressurized above atmos heric ressure or vor is being done vhich has the potential to drain hc vessel the pressure su pression oo r lcvc an empcrature shall be maintained vithin thc following limits.a.Minimum vater level~-6.25" (differential prcssure control>0 psid)-7.25" (0 psid differen-tial prcssure control)a.Sec Vusghuh'~n Q~cgae6 ts Pr gFg Zygo c Such('ca@on 6 r cnanges p, BFg Is~s 3.s.x e suppression chamber vater level bc checked once Whenever hea~P.s.g.q,t,) is added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall bc continually monitored and shall be observed and logged every 5 minutes until the heat addition is terminated. b.Mazimum vater level=lit AM 5'Ra.g.~,t.l f<s~frequency-~ce/s pp~,>Unit 1 3.7/4.7-1 PAGE W OP> 0 ~p<<il'ca8rn 3.L.Z./AU8 23 1991~~BcT(oA A Cond iHOn 8 f4+eel pg~',<eg Re%on c.ith the suppression pool p water temperature >95'F 3 initiate pool cooling and leo 3,Q,g,f QeR~~Q fk+iOn g,2 LI ACT(oe 8 I o<u'ref agon D,g estore the temperature o g 95 F vithin 4 hours or e n at least H OWH COHDITIOH vithin thc next 6 hours and in the COLD SHUTDOWH COHDITIOH vithin the folloving 30 hours.LI Alen pny os<A>'iE,+p~p->5/Vo Cia;s~op h,il~<+1<on gran e 7 54TIOpJ C d.Conlitjon Q Rca~red WvuN@el With the suppression pool vater temperature >105'F during testing of ECCS o relief valves, top all gCt 3.C.w.l.k LQ t test ng, it ate oo o follov thc t.act on in Specification g~$3.7.k.l.c above.OHDITIOH, HOT SThHDBY COHDITIOH (with all control rods not inserted), or REh OWER OPERATION the reactor s c Aa~o,i l scr P~e RCRaaio<gf +c+jo~g With the suppress'ion C~;tpool vatcr temperature 6'120'P o ov ng reactor isolation epressurize to<200 psig at normal'P coo rates.~lo/V E e.With thc suppression pool vater temperature )Cond,yn P>110~P c.co 3.6 0 1.4 rOPPli a&i I'+y Wches t g 4-g BFH Unit 1 3.7/4.7-2 AMENGMENT No.I 85 PAG'E UNIT 2 CURRENT TECHNICAL SP ECIF ICATION MARKUP

LI 3.7 0 S S 4.7 CO S S b Applies to the crating status f the primary secondary c tainmenH syst h lies to the pr ary and sec dary containme integ ty.be ve~0b~v To assur the integrity of e primary secondary containment ystems.To verify th integrity of the rimary and secondary containment. Aqui'eaQ,@ ht any time that the irradiated fuel is in the reactor vessel, and the nuclear system is pressurized above atmospheric pressure or wor s e ng one which has the potential to drain the vessel the pressure suppressi ater leve emperature shall be maintained within the following limits.a.Minimum water level=-6.25" (differential pressure control>0 paid)-7.25" (0 psid differen-tial pressure control)b.Maximum water level~1 tl~~s~~flclfP A g, C,HhM<~~its).S.Z em Sus TI FI c'ATI4N peg CHAN pcs t-o g.BFN>.6 z 2 a.The suppZession chamber water level be checked once er day.Whenever hea~SR36i~ I s added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall be observed and logged every 5 minutes until the heat addition is terminated. kid.sn s.a.~.l I 4"~4<y<WC~-n CA./Z'll l4 5 BFH Unit 2 3.7/4.7-1 PAGE A GF~

NQY 18]988>iopoSc4 (2C)~'rC4 Ac+un 4 t Cvloh/A QAClo~0~gpgur rQ pc.kid~A 2-4/QcTr o 8 gegwsr+)le id'.3 ith the suppression pool J vater temperature >95'F nitiate pool cooling and restore the temperature to g 95'F vithin 24 hours o be in at east t e HOT SHUTDOWH COHDITIOH vithin the next 6 hours and in the COLD SHUTDOWH COHDITIOH vithin the folloving 30 hours/ Ae4;o~e.i J4.Tio~q ith the suppression pool vater temperature >105'F during testing of ECCS or relief valves, stop all testing, t turbo 1 follov the action in Specification 3.7.h.l.c above.g.c.z.J 6 pc.Tla<4'>oN E e~7eg.g Aek~~b.i f.Co4<<.c E With the suppression pool vater tern erature>110'F during the STAR OHDITIOH, HOT STANDBY COHDITIOH (vith all control rods not inserted), or REACTOR WER OPERATIO the reactor shall be sc ed Pr o go~Pe u.t a8 A(4 o e D~With the suppression pool vater temperature 120'F folloving reactor isolation epressurize to<200 psig at norma oo ovn rates.geo g.c.2.l.c App),c,'Ll,~y Naos l~Z+5 Unit 2 3.7/4.7-2 AMENDMENT NO.y5 4 PAGE~OF 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP ~7 4.7 Appli to the operating status of the rimary and second containm t systems.Applies to the rimary and econdary conta ent tegrity>>To assure th integrity of the primary and s ondary tainment systems~To ver fy the integri of the primary secondary containment'. At any time that the irradiated fuel is in the pyi;(gb,f, p]reactor vessel, and the nuclear system is pressurized above atmospheric ressare r wor s e done which~the potential to dra the vessel the pressure ress on ol ster leve 1 e maintained within the'following limits.a.Ninimam water level~-6.25" (differential pressure control>0 paid)-7.25" (0 paid differen-tial pressure control)SCC$g5tgWh4+Foi Chan)t.'<&pe i5fb i S.w Yuc8A'colon Ai Chan)C'j 0 i 1<4 34,a x a.The suppression chamber water level be checked once er day.Whenever heaW>R w aLt s added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall be observed and logged every 5 minutes until the heat addition is terminated. b.Haxlanm water level~lit Afd sP.F.(.zeal.l Arsw FV<z~emy-once Qfpsurg BFE Unit 3 3.7/4.7-1 PAGE OF

SR.~;g~~Z.g.2.t NDV is Z88 ProgoLcd CCq~;rc Co c'.'~4>fun AA4Lsrg~A4tfon hL hon AL khaki(c With the suppression pool vater temperature >95 F tiate pool cooling, and restore the temperature to g 95'F vithin 24 hours r e at east the HOT SHOTDOW5 CONDITION vithin the next 6 hours and in the COLD SHUTDOWN CO%)ITI01 vithin the folio~30 hours Lco Z.6.2.l.a LI&pen cchy oPVRRO)E~m cWh~Q5/t~cf>u>sion 4I Bell Scolc~R~c 7+bio n C.d.ith the auppreaaicm 1 vater temperature 105'F during testing f HCCS or relief alvea top all Ro+~one'a'po o ov the testing, e e on in Specification 3.7.k.l.c above.~g,e.2.l.b]KHo Ipe4>n D Pea ircA~hc Yon p.l f.C~l Ken E With the suppression pool vater t eraturel>)104 ur'the S HDITI01, HOT STh2IDBY CONDITIONl,'vith all ccmtrol rods not inserted), or 0 the reactor shall be totogc c~,'c hm With the suppression pool vater temperature >1204F o lo reactor iaolaticm, epreaaurize o c 200 pai norma coo ovn rates.L~g,C.Z.l.c Afflr'CebrlS7 l'muon f"a+3 BF5 Unit 3 3,7/4.7-2 AMENDMENT NO.j2 9 PAGE'OF 3 0 JUSTIFICATION fOR CHANGES BFN ISTS 3.6.2.1-SUPPRESSION POOL AVERAGE TEHPERATURE ADH IN ISTRATI VE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no.technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.TECHNICAL CHANGES-NORE RESTRICTIVE ~M1 H2 H4 Existing Specification 3.7.A.l.e modifies the applicability governing suppression pool temperature such that the temperature limit applies only during the STARTUP CONDITION, HOT STANDBY CONDITION (with all control rods inserted), or REACTOR POWER OPERATION. Proposed LCO 3.6.2.1, Suppression Pool Average Temperature, ACTION D is applicable in Hodes 1, 2, and 3.Therefore, this change is more restrictive. CTS Surveillance Requirement 4.7.A.l.a only requires continual suppression pool temperature monitoring and logging whenever heat is added to the suppression pool during testing.Proposed SR 3.6.2.1.1 is more restrictive since it also requires this verification be performed once every 24 hours in the absence of testing.A new Required Action has been added (proposed Required Action A.l)to verify temperature is c 110'F every hour, anytime temperature has exceeded 95'F.This is an additional restriction on plant operation. When temperature exceeds 110'F, the current requirements only require the reactor to be scrammed.Proposed Required Action D.2 requires the temperature to be verified a 120'F every 30 minutes and a cooldown to HODE 4 within 36 hours, respectively. If temperature exceeds 120'F, the current requirements only require the RPV to be depressurized to<200 psig at normal cooldown rates.Proposed ACTION E now requires the 200 psig limit to be attained in 12 hours, and to continue cooling down the plant to cold shutdown (HODE 4)within 36 hours.These are additional restrictions on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.1-SUPPRESSION POOL AVERAGE TEMPERATURE The proposed ACTION (ACTION E)when pool temperature exceeds 120'F does not depend upon whether the reactor is isolated.If pool temperature reaches 120'F, regardless of whether the reactor is isolated, significant heat could still be added to the suppression pool and the Required Action is appropriate. Even with the reactor not isolated, there may be no heat rejection from the containment, as in the case of loss of condenser vacuum.Applying the actions regardless of whether the reactor is isolated does not introduce any operation which is unanalyzed. This change is more restrictive on plant operations. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1 Details of how to reduce suppression pool temperature to within the limits have been relocated to plant procedures. Methods for restoring pool temperature are more appropriately located in plant procedures. Changes to the procedure will be controlled by the licensee controlled programs."Specific" The Applicability for proposed LCO 3.6.2.1, Suppression Pool Average Temperature, is Modes 1, 2, and 3.However, this Applicability is modified within LCO 3.6.2.1 so that a lower suppression pool temperature limit applies if any Operable IRM channel is on Range 7 or above.This limit was selected so that the suppression pool temperature limits are applicable when the reactor is critical with reactor power approximately at the point of adding heat.As a result of this qualification to the Applicability statement, suppression pool temperature is required to be maintained at a temperature of less than 95'F (or less than 105'F while performing tests that add heat to the suppression pool)only when the reactor is critical with reactor power at the approximate level where heat generated is approximately equal to normal system heat losses.If the reactor is not critical or at a power below the point of adding heat, the suppression pool may be maintained at an average temperature up to 110'F.This change is less restrictive because CTS 3.7.A.1.required the lower suppression pool temperature to be less than 95'F (or less than 105'F while performing tests that add heat to the suppression pool)even if the reactor is not critical or not above the point of adding heat.If the reactor is not critical or the reactor is below the point of adding heat, there is significantly less heat generation from decay heat than assumed in the design basis.The suppression pool is designed to absorb the decay heat and sensible energy released during a reactor blowdown via safety/relief valves or from design basis accidents when the reactor has been operating continuously at full power for a BFN-UNITS 1, 2, 5 3 Revision 0 pAGE~OF~ 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.1-SUPPRESSION POOL AVERAGE TEMPERATURE considerable period of time.Any event initiated with reactor power or reactor power history less than these conditions will place considerably less heat load on the suppression pool than a DBA LOCA.This change is consistent with NUREG-1433. In addition, the shutdown requirements, if the temperature is not restored, have been modified to only require reducing power to below IRH Range 7 within 12 hours, consistent with the new Applicability. BFN-UNITS 1, 2, 5 3 Revision 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 4.7 ab Applic to the opc ating status of the imary and econdary containm t systems.A lies to thc p imary and se ndary contai t inte rity.~0i~~v To assure the tegrity o the primary and secondary ontainmcnt systems.To verify he integrity f the primary and ccondary ontainm A.1.At any time that the~)<cobe]july irradiated fuel is in the reactor vessel, and the nuclear system is pressurized above atmospheric pressure ng one vhich has thc potential to drain thc vessel e pressure suppress on pool vater level an tern cra ure shall bc maintained vithin t e folloving limits.a.Mininnnn vater level~-6.25" (differential pressure control>0 psid)-7.25" (0 psid differen-tial pressure control)a.JILS fjA~Oh Ar Ckg~~Wc Ph IsTS 3,5,2 See'u~h~qg>~A c Oa~~&r SF'S75 3, Q.2,I Thc suppression chamber vater level bc checked once per day cne r eat s added to thc suppression pool by testing of the ECCS or relief valves the ool temperature shall be continually monitored and shall be observed and logged every 5 minutes until the heat addition is terminat8d. b.Maximum vater level~1N LI fr'~gaioeR L~I~<4 heres s BFH Unit 1 3.7/4.7-1 PP,QE~OF

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 0 SE'C<csdifiqih~ ~c-P4i'.>4ira, (o C~4~Pt Q~$fhl (STT>.4,g.l A>Z4'.2.4.7 plies to the opc ting status of c primary and s condary coat iameat systems.0 To assure he iatcgrity of primary aa secondary containment stems.hppl s to the prima and scc ry containment integri Qhiee~~X To verify the tegrity of the primary and seco ary ontainmcnt. P~i;ops,'/Igy I.ht, any time that thc irradiated fuel is in the reactor vcsscl, and the nuclear system is pressurized above atmospheric prcssure or vork is being done vhich has the potential to drain the vessel e pressure suppression pool vater level erature hall be ma nta ned vithin t e folloviag limits.a.Minimum vater level=-6.25" (differential pressure control>0 psid)-7.25" (0 psid differen-tial pressure control)g(~SR 3.<,z.z.I ao SM MS TIFicATIod FoR CHAA WS Fog.LFN iSlX 3.5;2.3eE'g EST ifiCA7lohJ FoR Hhuo'es~g D&l isis 3.g z I Thc suppgession chamber vater level bc checked once per day.Whenever eat s a ded to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall be observed and logged every 5 minutes until the heat addition is terminated. b.Maximum vater level~1 tt')cScd AC.TiO+P, L~P opo5ed ACTioN 8 BFH Unit 2 3.7/4.7-1 PAGE~OF~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP pAGE OF 2, ace r~y~soo kv Chape Ate SftCS'ECCL 9%o 6e2~ni 3.7 4.7 n hpplie to the operating tates of, the p and second containm'aystcsls e hppliea to the imary and econdary conta ent egrity.To assure the egrity of the primary and sec containment syat To ver the integri of the primary secondary contalnm t.Ll:o 3'.4.2,..1 it any time that the irradiated fael la ln the reactor vessel, ancL the nuclear system ls preaanrixed above atmospheric pressure r vorR ia e one ch~the potential to draln the vessel e pressure reaaion ol eater level and t crater 1 be ma tained the'follovtng limits.a.Minimum vater level~-6.25" (differential pressure control>0 paid)-7.25" (0 paid differen-tial pressure control)~c'~4cahq+c~SF'sis 9.g.y.See aeggiaatjon 6c'harta&a bN iSv3'z.s,z. s added to the suppression pool by testing of the ECCS r relief valves the ol temperature shall be continually monitored and shall be observed and logged every 5 minutes until the hea addition la terminat Maahm SR Z,t,.z.z,t a.The suppression chamber vater level be checked once per day.whenever a b.Haxha~eater level~lit QLl~e'cct)kwon it c.W&coon 8 BFÃUnit 3 3.7/4.7-1 t JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.2-SUPPRESSION POOL WATER LEVEL ADNINI STRATI VE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be mor e readily readable, and therefore, understandable by plant operators as well*as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.TECHNICAL CHANGES-LESS RESTRICTIVE"Specific" The existing Action for suppression pool water level outside limits (Specification 3.7.A.1)allows no time to restore level.An unanticipated change in suppression pool level would require addressing the cause and aligning the appropriate system to raise or lower the pool level.These activities require some time to accomplish without undo haste.The out-of-service time is based on engineering judgement of the relative risks associated with: 1)the safety significance of the system;2)the probability of an event requiring the safety function of the system;and 3)the relative risks associated with the plant transient and potential challenge of safety systems experienced by requiring a plant shutdown.Upon further review, and discussion with the NRC Staff, during the development of the BWR Standard Technical Specifications, NUREG 1433, a 2 hour restoration allowance was determined to be appropriate. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.2-SUPPRESSION POOL MATER LEVEL L2 Per CTS, if suppression pool water level is not maintained within limits, the Specification is violated and in accordance with TS 1.0.C.l the plant must be placed in Hot Standby within 6 hours and in Cold Shutdown within the following 30 hours unless suppression pool water level is restored.This provides actions for circumstances not directly provided for in the specifications and where occurrence would violate the intent of the specification. The BFN ISTS provides Action within the Specification which could be considered less restrictive than CTS.Action B allows 12 hours to be in MODE 3 (Hot Shutdown)and 36 hours to be in MODE 4 (Cold Shutdown). The proposed Action is considered less restrictive since 12 hours is allowed to place the unit in Hot Shutdown versus the 6 hours allowed to place the unit in Hot Standby per CTS.BFN-UNITS 1, 2, 5 3 Revision 0 4 UNIT I CURRENT TECHNICAL SPECIFICATION MARKUP ra

ent 6oc6~~3.b>3 Thc RHRS shall be OPERABLE 8.(1)PRIOR TO STARTUP from a COLD CONDITION; or l.a.Simulated'utoma tic Ac tua tion Test Once Operating Cycle Rpfh mbi lily (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.b.C~d.e~Pump OPERA-Pcr BILITY'pecificatio 1.0.MM Motor Opera-Per ted'valve Specification OPERABILITY 1.0.MM Pump Flow Rate Once/3 months Test Check Pcr Valve Specification 1.O.MM SW GuS+IPicah'o~ Qgggg.kr gpH lsd Once/Mont Verify that each valve (manual,, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, is in its correct position.8~Verify LPCI Once/Month subsystem cross-tie valve is closed~power removed from valve operator.Low pressure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperablc. BFN Unit 1 3.5/4.5~Except that an automatic valve capable of auto-maticc return to i ts ECCS position when an ECCS signal is present may be in a position for another mode of operation. AMENDMENT NO.2 04 PAGP-k OP'I 4 .2.3 AUG 02 1989 Action/JC1 loni 8 If one RHR pump (containment cooling mode)or associated heat exchanger is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pumps (containment cooling mode)and associated heat exchanger and diesel enera ors d a access paths of the RHRS (containment cooling mode)are OPERABLE.See Sushi<'~hboA~ Ck~n)c5 W 8~hi I575 3,g,>6.If tvo RHR pumps (containment cooling mode)or associated heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mode), the a ciated heat exchangers iesel generate s all access pa of the RHRS (containment cooling mode)are OPERABLE.If tvo access paths of the RHRS (containment cooling mode)for each phase of the mo dr@we sup'pression chamber s rays and suppress on pool cool ng)are not OPERABLE, the unit may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the mode remains OPERABLE.(~p a t ona Sk'g s.z.z.i~R a.I..Z,p,> Se'<gus~'fi'cab'on CQ+~~f3PA ISIS 3.4.2.9+3.L,w.s t~oPos~d ACgtOg C BFH Unit l 3.5/4.5-6 0 8./l.77onl 847 s~p~sftlous~em~>and, If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be uninitiated and the reactor shall be laced in the COLD SHUTDOWN CONDITION within hours.Bc When the reac or vesse pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall bc OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low prcssure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwisc inoperable. ce 9.When the reactor vessel prcssure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification I.O.MM.10.If the conditions of Specification 3.5.A.5 are met, LPCI and containmcnt cooling are not required.10.No additional surveillance required.BPN Unit l.When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 belo~.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperablc if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall bc.demonstrated to be OPERABLE per Specification 1.0.MN when the cross-connect capability is required.+c Uus&&aago~ W C~g~~BF'9 ls rs 8.S.)y~~.~ANENOMENt'NO. 20 C UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP ent Coe+hzg+tainment 4t'0 3.Co iQ 3 1.The RHRS shall be OPERABLE (1)PRIOR TO STARTUP from a COLD CONDITION; or a.Simulated Once/Automatic Operating Actuation Cycle ApPI;~l till'2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.Z, through 3.5.B.7.b.C~d.e.Pump OPERA-Per BILITY Specification 1.0.MM Motor Opera-Per ted valve Specification OPERABILITY l.O.MM Pump Flow Once/3 Rate months Testable Per Check Specification Valve 1.0.MM See.Zuskikeuku~ Eo C~.~4sr SFM (s~3.g.(Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked sealed or otherwise secured in posi-tion, is in its correct position.f3 Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. gu Verify LPCI Once/Month subsystem cross-tie valve is closed~power removed from valve operator.Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. BPÃUnit 2 3.5/4.5-4 AMENOMENT KO.2 28 PAGE OF~ Al CooMn~g~AU6 02 58 S st ntainmcnt gQWiog p 5.If one RHR pump (containment cooling mode)or associated heat exchanger is inoperable, thc reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pumps (containment cooling mode)and associated heat exchangers diesc enerators all access of the RHRS (containment cooling mode)are OPERABLE.rctpl+pccL' 5ea~~s440;4('.~ 4~PM Is~5 B,.zl 6~If tvo RHR pumps (containment cooling mode)or associated heat exchangers are inoperable, thc reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cobling mode), thc associated heat cxcha11gersg csc generators access pa of the BHRS (containment cooling mode)are OPERABLE.(4L SQ 3~t'o.2.3.(sg.r.l..z.r.. z ce 7~If tvo access paths of the RHRS (containment cooling mode)for each phase of thc mode e sprays, suppression chamber s ra and suppression pool cooling)are not OPERABLE, the unit may remain in operation for a period not to cxcced 7 days provided at least one path for each phase of the mode remains OPERABLE.S<e T~s7-IF'tcATIafv pop cAAl4G-Eg Fyg~>>ebs Propose&A<T~ad c BFH Unit 2 3.s/4.s-6:ANENDMENNO. Ib 9 Cl 3.5.B ent 8.QC jlo+QO7>guest~+C4C Dln4td p~o)g.4~~9 If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed ia the OLD SHUTDOWN CONDITION within hours.J Z 96 When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. ce 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Speci,f ication 1.0.MM.0.If the conditions of Specification 3.5.A.5 arc met, LPCI and containment cooling are not required.10.No additional surveillance required.When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-conncct capability is not a short-term requirement, a component is not considered inoperablc i.f cross-connect capabi.lity can be restored to servi.ce within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be demonitrated to be OPERABLE per Specification 1.0.MM when the cross-connect capabi.li.ty is re uired.+7I t I c A7 (y p'Fbg.CHh U~S'~~~5.I 6 R.s.>AMENOMENT RU 2 23 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP SPzc, P 3 6.2.3 kyUca b;);yy (1)PRIOR TO STARTUP from a COLD CONDITION'r (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.LCa g.(.2+3 8.1.The RHRS shall be OPERABLE l.a.Simula ted Automa tie Ac tua t ion Test Once/Operating Cycle.b Pump.OPERA-Per BILITY Specification 1.0.MM d Pump Flow Rate Once/3 months e~Testable Check V a).ve Per Specification 1.0.MM c.Motor Opera-Per ,ted valve Specificatio OPERABILITY 1.0.MM~K 5%S'can o~"8>4'~Bl=e)spy 3,5 (Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, ig in its correct.position.Verify LPCI Once/Month subsystem cross-tie valve is closed~power removed from valve operator.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation'for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another ode of operation BFN Unit 3 3.5/4.5-4 em~~er e.z 77 PAGE~OF~ AUB 02 1988 5~BQjon P If one RHR pump (containmcnt cooling mode)or associated heat exchanger is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the rcmainiag RHR pumps (containment cooliag mode)aad associated heat exchanger diesel generators all access pa o the RHRS (contaiamcat cooling mode)are OPERABLE.Qpl~+54s M'cnMn+~~g&C SION ISIS g,g,t ocHo~8 6~I If tvo RHR pumps (containment cooliag mode)or associated heat cxchaagcrs are inoyerable, the reactor may remain in operation for a yeriod not to exceed 7 days provided the rcmainiag RHR yes (contaiamcnt cooling mode), the associated heat exchaag era ese g erators all access pa of the RHRS (containment cooling mode)are OPERABLE.aM sRz.r,.z.z, SR S.b,g.p.m 7~If tvo access paths of the RHRS (coataiament cooling mode)for each phase of the mode e 1 spra suppression chamber sprays aut suppress oa poo cooling)are not OPERABLE, the unit may remain in operation for a yeriod not to cxcecd 7 days provided at least onc path for each phase of the mode remains OPERABLE.5K~g~{ic~n QQ.Chgc5,~~~~)Sent Z.<.a.q m Z.e.a,s rope/QQo rl J BF5 Unit 3 3.5/4.5-6 NENMNTNO.SOO 3.5.B 8.Rh'on p HOT Stlut~d&4Dif>owl a4 lghe s (ontainm t If Specifications 3.5.B.1 through 3.5.B.7 are not met, orderly shutdown shall be initiated and the reactor shall bc lace in the S WN CONDITION~ithin hours.Z 3C, oo xng 8.inment nc 9.When the reac or vesse pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one" RHR loop<<ith two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually rcaligncd and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.0.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.10.No additional surveillance required.BFN Unit 3 When there is irradiated fuel in the reactor and the reactor is not in thc COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves.on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-conncct capability is not a short-term requirement, a component is not considered inoperablc if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The B and D RHR pumps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specif ication 1.0.MM when the cross-connect capability is required.5~4~usga'azfit)n 4(chgwgcc@'Pg tSq S s.S.I g3.g.~AMENDMENT NO.y.77 pp,GE or-JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.3-RHR SUPPRESSION POOL COOLING ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.TECHNICAL CHANGES-MORE RESTRICTIVE t Ml Surveillance Requirements (SR 3.6.2.3.1 and 3.6.2.3.2) have been added to ensure that the correct valve lineup for the RHR suppression pool cooling subsystems is maintained and RHR pump testing is performed to ensure the RHR suppression pool cooling subsystems remain capable of providing the overall DBA suppression pool cooling requirement. This change is consistent with NUREG-1433. M2 CTS 3.5.B.8 requires an orderly shutdown be initiated.and the reactor to be in the Cold Shutdown Condition within 24 hours when required RHR suppression pool cooling subsystems are inoperable. Proposed Action 0 will require the plant be in MODE 3 (Hot Shutdown Condition) in 12 hours and MODE 4 (Cold Shutdown Condition) in 36 hours.The addition of this intermediate step to the Cold Shutdown Condition is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.BFN-UNITS 1, 2, 5 3 Revision 0 Cl JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.3-RHR SUPPRESSION POOL COOLING TECHNICAL CHANGES-LESS RESTRICTIVE Ll Proposed ACTION C will allow 8 hours to restore required RHR suppression pool cooling subsystems to operable status prior to initiating a shutdown.The proposed 8 hour Completion Time provides some time to restore the required subsystems to Operable status, yet is short enough that operating an additional 8 hours is not risk significant. Only 8 hours is allowed since their is a substantial loss'f the'primary containment bypass leakage mitigation function.The 8 hour restoration time is considered acceptable due to the low probability of a DBA and because alternative methods to remove decay heat from the primary containment are still available. In addition, if the required subsystem(s) are restored to Operable status prior to the expiration of the 8 hours, a unit shutdown is averted.Thus, the potential of a unit scram occur ring while shutting the unit down, which then could result in a need for a subsystem when it is inoperable, has been decreased. L2 The time to reach NODE 4, Cold Shutdown, has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in NODE 3 (Hot Sh'utdown) within 12 hours has been added.These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1;2, 5 3 Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP pAGE E III ~/'f licobilil-) (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.~ockk~LCO ZiC~2ig 8.l 1.The RHRS shall be OPERABLE (1)PRIOR TO STARTUP from a COLD CONDITION; or I.a.b.C~d~nt Simula ted Automatic Ac tua tion Tes t Once/Operating Cycle Pump OPERA-Per BILITY Specification 1.0.MM Pump Flow Once/3 Rate months Test Check Per Valve Specification 1.0.MM Motor Opera-Per ted'valve Specification OPERABILITY 1.0.MM 5c'e SK5+0'+on Q~ggc 8PIJ lsd 3,g, J Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, is in its correct position.Verify LPCI Once/Month subsystem cross-tie valve is closed gaul power removed from valve operator.BFN Unit 1 Low-pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5-4*Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. AMNOuENT HO.20'ATE pp 5

AUG 02 tggg 3.5.B ova S t~g~(LPCI and Containment Cooling)4.5.B.s ua ca e ova S st Qg~(LPCI and Containment Cooling)4.5.B.1 (cont'd)2.With the reactor vessel pressure less than 105 psig, the RHRS may be removed from service (except that tvo RHR pumps-containment cooling mode and associated heat exchangers must remain OPERABLE)for a period not to excccd 24 hours vhile being drained of suppression chamber quality vater and filled vith primary coolant quality vater provided that during cooldown tvo loops vith one pump per loop or one loop vith tvo pumps, and associated diesel generators, in the core spray system arc OPERABLE.Each LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Tvo LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system pressure of 250 sig.2.SR 3'L,z.g,z LR)Se<3~qqi~~~g, g~~~~B~e icy',<.z.S An r tes on the ryvel and torus headers an nozzles shall be conducted once/5 years.v ter tes c form on thc oru he cr in ieu o th air st.3.If one RHR pump (LPCI mode)is inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (LPCI mode)and both access paths of thc RHRS (LPCI mode)and thc CSS and thc diesel generators remain OPERABLE.3.Ho additional surveillance required.4.If mqr 2 RHR pumps (LPCI mode)become inoperable, thc reactor shall be placed in the COLD SHDTDOWH COHDITIOH vithin 24 hours.4.Ho additional surveillance required.Sc~Z~sw~Bee~sag BPH Unit 1 3.5/4.5-5 PAGE~~'P-'" QIENOMENT NO.16 9 QeciQrcqg~ p.g,2 y AUG 02 t989 5.Bono/V If one RHR pump (containment cooling mode)or associated heat exchanger is inoperablc, the reactor may remain in operation for a period not.to exceed 30 days provided the remaining RHR pumps (containment cooling mode)and associated hea cxchangers and diese enera or and all access paths of the RHRS (containment cooling mode)are OPERABLE.<<<a SWAah'C~f r Qppf)5y5 3 8'l.n~e 8 6.If tvo RHR pumps (containment cooling mode)or associated heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR yumys (containment cooling mode), the associated heat exchangers, iese enera and all access pa of the RHRS (containmcnt cooling mode)are OPERABLE.Add s'g g.7~If tvo access paths of the RERS (containmcnt cooling mode)for each phase of the mode dryve prays ress on er sprays, sad ress on oo cooling am not E LE, thc un msp'emain in operation for a period not to exceed 7 days provided at least one path for each yhase of thc mode remains OPERABLE.Sec y<s+5(~agon Sac Changes 4~N=o I5T's, v.a.~.3+3,(,z,5 LI~&58k+T (pg Q BFE Unit 1 3.5/4.5-6 NENDMENTNO. 16 9

nt nment 8.kTJ0N D 9.If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SH WN CONDITION within ho s.C.2, When ac or vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.Thc pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection'LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. ~a~.Nz>n+he QT SHQYSOM&no>p'ion))g When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Speci.fication 1.0.MM.0.If the condi.tions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.10.No additional surveillance required.BFN Unit 1 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION,' RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect apability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours'~)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MN when the cross-connect capability is required.5<<5'w'0'ca4og 0<5FQ ISTs z.s,llI g>~'AMENDMENT NO.2 0 4 UNIT 2 CURRENT TECHNICAL SP ECIFICATION MARKUP 3.5.B on ax ent 4.5.B nt~iraq~LCo$.42.Lt 1.The RHRS shall be OPERABLE (1)PRIOR TO STARTUP from a COLD CONDITION'r Bc'?b~)1.a.Simulated Automatic Actuation Test Once/Operating Cycle a.gpss;~4 l;4p (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.b.Co d.e.Pump OPERA-Per BILITY Specification 1.0.MM Motor Opera-Per ted valve Specif ication OPERABILITY 1.0.MM Pump Flow Once/3 Rate months Testable Per Check Specification Valve 1.0.MM See.xuSC(f'~]4 g,~.I%~EFW lSrr 3.S: I Once/Month Verify that each valve (manual, power>>operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi tion, is in its correct position.ge Verify LPCI Once/Month subsystem cross-tie valve is closed~power removed from valve operator.BFN Unit 2 3.5/4.5~Low prcssure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another ode of oper tio AMENOMBfT Ro.2 2 S

AUG 02 t88g t 3.5.B R du v S te~~(LPCI and Containment Cooling)4.5.B.s ua at e ova S t Q@ESS.(LPCZ and Containment Cooling)4.5.B.1 (cont'd)2.With the reactor vessel pressure less than 105 psig, the RHRS.may be removed from service (except that tvo RHR pumps-containment cooling mode and associated heat cxchangers must remain OPERABLE)for a period not ,to exceed 24 hours vhile being drained of suppression chamber quality vater and filled vith'rimary coalant quality water provided that during caoldovn tvo laops vith one pump per loop or one loop with tvo pumps, and associated diesel generators, in the core spray system are OPEUSLZ.Each LPCZ pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Tvo LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system pressure of 250 psig.2.An t s an th dryvel Sg~~>~>an torus headers an nozzles shall be conducted once/5 years.water test may e per cd on 6 total heads in lie of the Sir te 3.Zf one RHR pump (LPCI mode)is inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (LPCI mode)and both access paths of the RHRS (LPCZ mode)and the CSS and the diesel generators remain OPERABLE 3.o a onal surveillance required.4.If any 2 RHR pumps (LPCI mode)bccomc inoperable, the reactor shall be placed in the COLD SHDTDOWH COHDZTIOH vithin 24 hours.4.Ho additional survcillancc required.Qj?48$/'licollpa gi g4~<+NFL ICW 3.g.(3.5/4.5-5 AMENDMENT NO.16 9 Al S Ac<(ad If one RHR pump (containment cooling mode)or associated heat ezchanger is inoperable, the reactor may remain in operation for a period not to ezceed 30 days provided the reams~RHR pumps (containment cooling mode)and associated heat exchangers and diese generators and all access paths of the RHRS (containment cooling mode)are OPERhBLE.rgsp~.+~~t g4 0(cqfi 0 n 4r 40(Ep/LJ/5~Ac%(od&If two RHR pumps (containment cooling mode)or associated 'eat exc8xangers are inoperable, the reactor map'emain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mods)~the associated heat exchangers, diesel enerato s aQ access paths of the RHRS (containment cooling mode)are OPEELBLE.~~~Ado', SR'.C.a.y/P" (70 If two access paths of the RHRS (containment cooling mode)for each hase of the mode dzpvell sprays suppress on chamber sprays, ress on are not OPERhBLE, the unit may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the made remains OPZMBLE.5&V~$7 IFJCPSTsdAJ PbR CHAhlk~<<~BFN/sr',@p 3 pg gg,~P~q~sM Ace>OW~BF5 Unit 2 3 5/4.S-6 PAGE:AM@WarrNO. re e inment~A)inment e.A('(ohl L.Z If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed n e COLD SHUTDOWN CONDITION within hours$4 ac or vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall bc OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low prcssure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. do pl,e Po~S(lu77N~hl CO~O(7 Sahl~Q(B~rS W~d 9.When the reactor vesse pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Speci.fication 1.0.MM.0.If the conditions of Specification 3.5.A.S are met, LPCI and containment cooling are not required.10.No additional surveillance required.BFN When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capabi.lity can be restored to service within 5 hours.).5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall bc demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.See v.us AC cako m ('~~, 8(c'hl IS~3'g (g gS.Z.AMENDMEMT m.2 23 0' UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Cl ent 4.5.B ent~CO 3 o 4 o 1.o<l 0.1.The RHRS shall be OPERABLE (1)PRIOR TO STARTUP from a COLD CONDITION'r l.a.Simulated Automatic Ac tuation Tes t Once/Operating Cycle~PfLECab,'f.Q (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Speci.fications 3.5.B.2, hrough 3.5.B.7.b.C~d.e~Motor Opera-Per ted valve Specification OPERABILITY 1.0.MM Pump Flow Rate Once/3 months Testable Check Valve Per Specification l.O.MM Pump OPERA-Per BILITY Specification 1.0.MM~SwS+Ac4un 4<8 apl)5 TS y, g)Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, ig in its correct position.go Verify LPCI Once/Month subsystem cross-tie valve is closed~power removed from valve operator.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment'nd operation'for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. BFN Unit 3 3.5/4.5-4 N@DMENr go;I 77 '.(o.2~9 AUB 02 I88S 4.5.B 4.5.B.1 (cont'd)2.Mith the reactor vessel pressure less than 105 psig, the RHRS may be removed from service (except that tvo RHR pumps-containment cooling mode aad associated heat cxchangers must remain OPE)ULBLE) for a period not to exceed 24 hours vhilc being drained of suppression chamber quality vater and filled vith primary coolant quality vater provided that during cooldovn tvo loops with one pump per loop or one loop vith tvo pumps, and associated diesel generators, ia the core spray system are OPERkBIS.3i If oae RHR pump (LPCI mode)is inoperable, the reactor may remain ia operation for a period not to exceed 7 days provided the rcmaixdLag RHR pumps (LPCI mode)aad both access paths of the RHRS (LPCI mole)and the CSS and the diesel generators rcmaia OPERhBLE o Each LPCI pump shall deliver 9000 gpa against an indicated system pressure of 125 psig.Two LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system pressure of 250 psig.2.hn ir test on thc cl SRS.c.z.i torus eadcrs and nozzle shall be conducted once/5 years.k v st map c erform oa t h$adc lieu th air st.Sc<Eussy'iF:iW'r r+rihtsys44 BR'STS Zi4 eX,5 3.Eo additional surveillance required.If any 2 RHR pumps (LPCI Node)become inoperable, the reactor shall be placed in the COLD SHOTDOMH COHDITION within 24 hours.4.Eo additional surveillance required.5<8 34s~f:~~'o<4 Ch~s 4r B~g lS~3.5 I BFS Unit 3 3.5/4.5-5)~*r~"XC 5 NENDMENTNO. ~a o

n 3.4.2.AUB 0 2 1988 4.5 B.5.If one RHR pump (containmcnt cooling mole)or associated heat exchanger is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pumps (containment. cooling mode)and associated he cxchcuwcr and diesel generators and all access pa of the RHRS (containment cooling mode)e OPEBkBIS 5.Sc'<Du s~Pea h'o n Qr C~~Ai 8PN lST5 P.y.i 4+n 8 6~7~If tvo RHR pumps (containmcnt c'aoling mode)or assaciatcd heat cxchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the rcmaQdng RHR pumps (containmcnt cooling mode), the associated heat cxchangera ese genera ors, and all access pa o e RHRS (cantainmcnt cooling mode)are OPElhLBLS. If tvo access paths of the RHRS (cantainmcnt cooling made)for each e of the a+de ell a r ression chamber sprays, ress on oo o are nat OPERhBLE, the unit may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the mode remains OPERABLE.Wt Rdd sR s.c,z,q,i'(g c~3wswf'azgo~g,(g~~~ISIS g,g,Z,~~><z>f~w~z Zy,c pAcE OF-5 BF5 Unit 3 3'/4'-6 AMMMENTNO. 1o o

3.5.B 4.5.B 8-If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be.initiated and th eactor shall be place in the COLD SHUTDOWN CONDITION within 24 hours.3 When actor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation.for shutdown cooling, if capable of being manually realigned and not otherwise inoperablc. ~kTCtk o fn He Q)QSHuygavp4 Co+DiTie I h IP boating AhD 9.When the reactor vessel pressure is atmospheric the RHR pumps and valve that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specif ication 1.0.MM.0.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling arc not required.10.No additional surveillance required.BPN Unit 3 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect.capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.3.5/4.5-7 11'.The B and D RHR pumps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.~ELVA'PIcgh'oQ gr gfel~cf gl ggjg]$'fg AMENOMEHT MO.I 77 PAGE~Ci t JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.4-RHR SUPPRESSION POOL SPRAY ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.TECHNICAL CHANGES-NORE RESTRICTIVE H2 Surveillance Requirement (SR 3.6.2.4.1) has been added to ensure that the correct valve lineup for the RHR suppression pool spray subsystems is maintained. This ensures that the RHR suppression pool spray subsystems remain capable of providing the overall DBA suppression pool spray requirement. This change is consistent with NUREG-1433. CTS 3.5.B.8 requires an orderly shutdown be initiated and the reactor to be in the Cold Shutdown Condition within 24 hours when required RHR suppression pool spray subsystems are inoperable. Proposed Action 0 will require the plant be in NODE 3 (Hot Shutdown Condition) in 12 hours and MODE 4 (Cold Shutdown Condition) in 36 hours.The addition of this intermediate step to the Cold Shutdown Condition is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.BFN-UNITS 1, 2, 8L 3 Revision 0 0 0 JUSTIFICATION FOR CHANGES'FN ISTS 3.6.2.4-RHR SUPPRESSION POOL SPRAY TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1 Details of the methods of, performing surveillance test requirements have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll Proposed ACTION C will allow 8 hours to restore required RHR suppression pool spray subsystems to operable status prior to initiating a shutdown.The proposed 8 hour Completion Time provides some time to restore the required subsystems to Operable status, yet is short enough that operating an additional 8 hours is not risk significant. Only 8 hours is allowed since their is a substantial'oss of the primary containment bypass leakage mitigation function.The 8 hour restoration time is considered acceptable due to the low probability of a DBA and because alternative methods to remove decay heat from the primary containment are still available. In addition, if the required subsystem(s) are restored to Operable status prior to the expiration of the 8 hours, a unit shutdown is averted.Thus, the potential of a unit scram occur ring while shutting the unit down, which then could result in a need for a subsystem when it is inoperable, has been decreased. L2 The time to reach NODE 4, Cold Shutdown has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in NODE 3 (Hot Shutdown)within 12 hours has been added.These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS, 1, 2, 5 3 Revision 0 PAGE~OF~ 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 inment~~an ent 1.The RHRS shall be OPERABLE 8.(1)PRIOR TO STARTUP from a COLD CONDITION or~ao Simulated'utomatic Actuation Test Once/Operating Cycle~lfl'Qh;l,g (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.b.C~d~e.Pump OPERA-Per BILITY Specification 1.0.MM Motor Opera-Per ted'valve Specification OPERABILITY 1.0.MM Pump Flow Once/3 Rate months Test Check Per Valve Specification 1.O.MM 0 5'ee T~s&l~on 4r~'<35,(Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection f 1ow-path that is not locked, sealed, or otherwise secured in posi>>'tion is ln its correct position.ge Verify LPCI Once/Month subsystem cross-tie valve is closed azLd power removed from valve operator.BFN Unit 1 Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5-4 E xcept that an automatic valve l capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. AMENDMENT go.2 P g PAGE 0 S c.4icn AUG 02 1989 3.5.B ova S st Qg~(LPCI and Containment Cooling)4.5.B.s dua ea e ova S tern Qgg~(LPCI and Containment Cooling)4.5.B.1 (cont'd)2.Mith thc reactor vessel pressure less than 105 psig, the RHRS may bc removed from service (except that tvo~RHR pumps-containment cooling mode and, associated heat exchangers must remain OPERABLE)for a period not to exceed 24 hours vhile being drained of suppression chamber quality vatcr and filled vith primary coolant quality vater provided that during cooldovn tvo loops vith one pump per loop or one loop vith tvo pumps, and associated diesel generators, in the core spray system are OPERABLE.Each LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Tvo LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system prcssure of 250 psig.2.An Sc a'.a.g,g.g r e on the drywell orus headers and nozzles e conducted once/5~~~~Q~Q'un Po r conge 4,i gl-~IS~s years.vater test may be er ormed on the torus header.in lieu of the air test.3.If one RHR pump (LPCI mode)is inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RER pumps (LPCI mode)and both access paths of the RHRS (LPCI mode)and the CSS and the diesel generators remain OPERABLE.3.Ho additional surveillanc r cquir ed.4.If any 2 RHR pumps (LPCI mode)become inoperable, thc reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.4.Ho additional surveillanc required.See@wc.+4;cat <~*, g~+$<<lSTS S.S,i BFH Unit 1 3.5/4.5-5 AMENDMENT NO.16 9

~kwon 9 g,2~AUG 02$989 ent~oIXng)te ainment i9C7lorJ 5.If one RHR pump (containment cooling mode)or associated heat exchanger is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pumps (containment cooling mode)and associated heat exchanger ese enerator and all access paths of the RHRS (containment cooling mode)are OPERABLE.Qg).o a+~3~s+<~ah'on 4 cgqcs b'<6 lsT$3.LI ance 6.7.If tvo RHR pumps (containment cooling mode)or associated heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mode), the asso ted heat exchaagers, d ener and all access paths of the RHRS (containment cooling mode)are OPERABLE.If tvo access paths of the RHRS (containment cooling mode)for each phase of the mode (dryvell sprays, uppress on e and suppression pool cooling)~are no PERABLE, e t may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the mode remains OPERABLE.QAi Rg sa s,e, z, s.~fpopgrA RTI o~<BFH Unit 1 3'/4.5-6 hMENONENT NO.16 9

nment 8.If Specifications 3.5.B.1 through 3.5.B.7 are not mct, an orderly shutdown shall be~initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within~~hours.3S in~hC H r SHOTOO~N in 12hrs An/lance 9.When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least onc RHR loop with two pumps or two loops with one pump per loop shall bc OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operat'ion for shutdown cooling, i.f capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, thc RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABL pcr Specification 1.0.MM.0.If thc conditions of Specification 3.5.A.5 are met, LPCI and containmcnt cooling are not required.10.No additional surveillance required.BFN Unit 1 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must bc OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.5 CuZlkt tip'so%Iud lsi QAwilh W GFW lST'S 3'S.l+'3,5.p.AMENDMENT NO.2 0 4 pAGE

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 5 s.ci4icathien 3.Q.2.g ent ainment 1.The RHRS shall be OPERABLE (1)PRIOR TO STARTUP from a COLD CONDITION; or (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.l.a.b.C~d.e.Simulated Automatic Ac tua tion Test Pump OPERA-BILITY Once/Operating Cycle Per Specification l.O.MM Pump Flow Rate Once/3 months Testable Check Valve Per Specification 1.0.MM Motor Opera-Per ted valve Specification OPERABILITY l.O.MM Once/Month Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, is in its correct position.f$Low pressure coolant injection (LPCI)may be considered OPERABLEduring alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Verify LPCI Once/Month subsystem cross-tie valve is closed gaul power removed from valve operator.Except that an automatic valve capable of auto-metic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. BFN Unit 2 3.5/4.5-4 AMENDMENT RO.2 2 S

OH 5 sf~ceAon QC,25 AUG 02 t98G 3.5.B KHRQ, (LPCI and Containmcnt Cooling)4.5.B.u Rcmova S st QQ~S (LPC1 and Containment Cooling)2.Pith the reactor vessel pressure less than 105 psig, the RHRS may be removed from service (except that tvo BHR pumps-containment cooling mode and associated heat cxchangers must remain 3PERABLE)for a period not to exceed 24 hours vhile being drained of suppression chamber qaality.vater and filled vith primary coolant quality vater yrovidcd that daring"ooldowa tvo loops vith one pump per loop or one loop vith tvo yamps, and associated diesel generators, ia the core spray system are OPERABLE.5'R S.C.2.S'.2. 2.Aa n the dryvell and torus headers and nozzles s e conducted once/5 years.vater test may b per ormed on the torus header in lieu of the air.test.A'e WdA(Cubo 4" C~Q>S 4r BF~is&RC.2.$4.5.B.1 (cont'd)Each LPCI pump shall deliver-9000 gpm against an indicated system pressure of 125 psig.Tvo LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system essare of 250 sig.3.If one RHR pump (LPCI mode)is inoperablc, the reactor may remain in operation for a period not to exceed?days provided the rcmaiaiag RHR pampa (LPCI mode)aad both access paths of the RHRS (LPCI mode)and the CSS and the diesel generators remain OPERABLE 3.Ho additional surveillance required.4.If any 2 RHR pamys (LPCI mode)become inoperable, the reactor shall be placed in the COXu SHOTBOmr COHDITIOH vithin 24 hours.4.Ho additional surveillance required.SIP, WR+fscRL~0o~~+5'tS BI hl ISTIC 3.Q,/BFH Unit 2 3.5/4.5-5 PAGE~OF~AMEMMENTNO. I,6 9 Pl 4~~a Sst aiamcnt g(.Z(oa A 5~If one RHR pamy (containment cooling mode)or associated heat cxchaagcr is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the renmixxiag RHR yamys (contaiamcnt cooling mode)and associated heat cx ers and diesel cnerato s access paths of the RHRS (containment cooling mode)are OPERABLE.6~If tvo RHR yamys (contaiamcnt cooling mode)or associated 'eat exchaagers are inoperable, the reactor map'emda in oyeration for a period aot to exceed 7 days yrovided the remaining RHR pampa (ccmtaiameat cooling mode), the associated heat dies cnerator 1 access yaths of the RHRS (contaiameat cooling mode)are OPERABLE.0 s 70 If two access paths of the RHRS (containmcat cooliag made)for each phase of the mode (dryvell syra press cm er syrays, and suppression pool cooling are not OPERABIZ, the t aalu rennin ia oyeration for a period aot to exceed 7 days provided at least one path for each yhase of the mode remains OPERABLE.+<sf<$~0jflow pop QLgr~~~40(gp'hl/57-g 4W 9.d.g.gf TroposM ALTiod C, BFE Unit 2 3 5/4.5-6:AMNnMan xO.ze 9 aa n aiament Qgi COUTin~g ainmen t 8.L2.If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall bc placed ia t e COLD SH CONDITION within ours.ea e reactor vessel pressure is atmospheric aad irradiated fuel is in the reactor vessel, at least onc RHR loop with two pumps or two loops with one pump per loop" shall be OPERABLE.Thc pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. QZ'~+4<-8+T$A'ergot N'~~Z 4c~w~cP 9.When thc reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to bc OPERABLE per Specification 1.0.MM.O If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.10.No additional surveillance required.BFN Unit 2 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on aa adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-conaect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to icrvice within 5 hours.)3.5/4.5-7 ll.The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to bc OPERABLE per Specification 1.0.MM when thc crose-connect capability is required.S~sVIF'~~r>c4 FoA cls~~ayr Pdk sf'<Ism 3 5 (pg g Z AMENDMENT HU.2 23

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Cl 0 0 3.5.B ent+1'cab;f;Q 1.The RHRS shall be OPERABLE 0.(1)PRIOR TO STARTUP from a COLD CONDITION; or (2)when there is irradiated fuel in the reactor vessel'nd when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.~a~b.C~d.e~Simula ted Automatic Actuation Test Once/Operating Cycle Motor Opera-Per ted valve Specificatio OPERABILITY 1.0.MM Pump Flow Rate Once/3 months Testable Check Valve Per Specification 1.0.MM Pump OPERA-Per BILITY Specification 1.0.MM ew wu>e,lsd%n*r chn~er 4u.~iSV5 g,g, I Once/Month Veri,fy that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, ig in its correct position.Veri fy LPCI Once/Month subsystem cross-tic valve is closed~power removed from valve operator.Low prcssure coolant injection (LPCI)may be considered OPERABLE during alignment and operation'for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. BFN Unit 3 3.5/4.5-4 hIENOMEe NO;I 77 pAGE~oF~ 0 5~i'ca5o~7 6 2.5 AUG 02 8y 3.5.B v S Q',ggQ (LPCl and Containmcnt Cooling)4.5.B.Qg~(LPCI and Containment Cooling)4.5.B.1 (cont'd)2~With the reactor vessel pressure less than 105 psig, the BHBS may be removed from service (except that tvo BHR pamys-containment cooling mode and associated heat exchangcrs mast remain OPEBASLE)for a period not to cxcecd 24 hours vhile being drained of suppression chamber quality vater and filled vith primary coolant quality vatcr provided that during cooldown two loops vith one pump per looy or one loop with two pumps, and associated diesel generators, in the core spray system are OPERhBLE.5R 34,2e5,Z Each LPCI pump shall dclivcr 9000 gpa against an indicated system prcssure of 125 psig.Two LPCI pumps in the same looy shall deliver 12000 gym against an indicated system yressurc of 250 psig.2.kn r tits on the dryvcll hcadcrs and nozzles shall bc conducted once/5 years.h vatcr test may be erformed on thc torus headc in lieu of the air test Sce XggC:caliban 4,~as g, g~H ISTIC><4 ex+3~If one RHR pump (LPCI mode)is inoyerable, the reactor may remain in operation for a period not to exceed 7 days provided the remanding RHR yamps (LPCI mode)and both access paths of the RHBS (LPCI mode)and the CSS and the diesel generators remain OPERhBLE.3.Eo additional surveillance required.4, If any 2 RHR pumps (LPCI mode)become inoperable, the reactor shall be placed in the COLD SHOTDOWH COHDITIOH within 24'ours.4.Eo additional surveillance required.5Cc ggcHQ~91=A$75 g,g,f BFH Unit 3 3.5/4.5-5 PAGE ANENDMENTNO. ~a o

U9 02 1989 4.5 B 5.4cTioe R If ane RHR pump (containmcnt cooling mode)or associated heat exchanger is inopcrablc, the reactor may remain in operation for a period not to exceed 30 days yrovidcd the remaining RHR yumps (containment caoling mode)and associated heat exchanger generators and access pa of the RHRS (containment cooling mode)are OPERABLE.5qc Wc+Ciech'on+, gyg~g go~Bw>lsd',f.i 6.If tvo RHR pumps (containmcnt cooling mode)or associated heat cxchangers are inoperable, the reactor may remain in operation for a period not to'exceed 7 days provided the remains~RHR pumps (cantainmcnt cooling mode), the assoc ed heat ers iese cnerators and all access of the RHRS (containment cooling made)are OPERABLE.Cegakredv I Ac@51 3 6eZ S:/7~If too access paths of the RHRS (containmcnt cooling mode)for each phase of the mode ell a rays, ression chamber sprays, and suppression yool cooling)ars not e t aalu remain in operation for a period not to exceed 7 days provided at least one path far each phase of the mode remains OPERABLE.c<tusw~Son Ar Qn~W BF~XSTS 9,6.Z.D~g 3.6.Z.qBHf Unit 3 scl R&bhl c 3.5/4.5-6 PAGE OF'IENMENTNO. la o

axnm nt ent 9.If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in COLD S WN CONDITION within hours.gk When t e reactor vessel pressure is atmospheric and.irradiated fuel is in the reactor vessel, at least onc RHR loop with two pumps or two loops with one pump per loop shall bc OPERABLE.The pumps'ssociated dicscl generators must also bc OPERABLE.Low pressure coolant injection (LPCI)may bc considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. <h+t.%g Sgk%~4 Andon i n I 2~rs nng 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Spccif ication 1.0.MM.0.If thc conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.10.No additional surveillance required.BEN Unit 3 When there is irradiated fuel ia thc reactor and the reactor is.not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangcrs, and valves'on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 ll.The B and D RHR pumps on unit 2 which supply cross-connect capability shall bc demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.gee~RPcahsn 4r C~cs~" S+<~5TS r.S.!+V.S> AMENDMENT ND.Z 77

t 53USTIFICATION FOR CHANGES BFN ISTS 3.6.2.5-RHR DRYWELL SPRAY ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no.technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.TECHNICAL CHANGES-MORE RESTRICTIVE Ml M2 L Surveillance Requirement (SR 3.6.2.5.1)has been added to ensure that the correct valve lineup for the RHR drywell spray subsystems is maintained. This ensures that the RHR drywell spray subsystems remain capable of providing the overall DBA drywell spray requirement. This change is consistent with NUREG-1433. CTS 3.5.B.8 requires an orderly shutdown be initiated and the reactor to be in the Cold Shutdown Condition within 24 hours when required RHR drywell spray subsystems are inoperable. Proposed Action 0 will require the plant be in MODE 3 (Hot Shutdown Condition) in 12 hours and MODE 4 (Cold Shutdown Condition) in 36 hours.The addition of this intermediate step to the Cold Shutdown Condition is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completioo Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF~

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.5-RHR DRYWELL SPRAY TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Details of the methods of performing surveillance test requirements have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll Proposed ACTION C will allow 8 hours to restore required RHR drywell cooling subsystems to operable status prior to initiating a shutdown.The proposed 8 hour Completion Time provides some time to restore the required subsystems to Operable status, yet is short enough that operating an additional 8 hours is not risk significant. Only 8 hours is allowed since their loss substantially reduces the ability to maintain primary containment within design limits.The 8 hour restoration time is considered acceptable due to the low probability of a DBA and because alternative methods to remove decay heat from the primary containment are still available. In addition, if the required subsystem(s) are restored to Operable status prior to the expiration of the 8 hours, a unit shutdown is averted.Thus, the potential of a unit scram occurring while shutting the unit down, which then could result in a need for a subsystem when it is inoperable, has been decreased. L2 The time to reach NODE 4, Cold Shutdown has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.Thisextra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in NODE 3 (Hot Shutdown)within 12 hours has been added.These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF 3,

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP

a.Di erent'al pressure betveen the dwell ancL suppression LCO~be shall be naincainecf 3 gg g at equal to or greater~psiC except as spec'ecf (1)and (2)belov: a.The pressure ct ferential betveen the Crjvell ant sayyression c'"amber sha'be" cl c'ast once each akmh~I 2 hots<~R2.()Ms CI e"mc'al shal'e establish'ithin 24 hours a eviag oyerat~tcRQeratur iCCj4>Irj r cesar e The cU.ff erencial pressure nay be raducecL to 1 ss than 1.1 s 4 hours r.or to a scheclalei bntdovn.Qo Now (2)This Cifferential nay be ctecrease4 co less than 1.1 ysicL for a aaxfmaa of four hours Carhg requireL oyerability teston of the HPCI systea, RCIC systea auf the 4ryvell-yressare sayyression chaaber vacuaI breakers.b If the differential pressure of Specification 3.7.4.6.a QgT fog5 cannot be maintainecL and the@if ermcial pressure cannoc+~be restorers vithin the subsecLaenc ~ho4,.It'ho~~Re4~c<7lfMnlAL/buaR fn lghrs BPH Unit 1 (l 3.7/4.7 12 PACE

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

S aiS Dr i a a.Differential pressure betveen LCO the dryvell and suppression chamber shall be maintained at equal to or greater than 1.1 psid except as spec'fied in (1)and (2)belov: Sg Q.io.2.4.a.e pressure dS ferential betveen the dryvell and suppression chambe shall be recorded at least anc ach~.Ia 4<<~(1)This differential shall be established vithin 24 hours f a eving~PP'4"'~P perat ng tern erature The differential pressure may be reduced to less than 1.1 sid 24 ours r or to a schedu utdovn (2)This differential may be decreased to less LCO than 1.1 psid for a bloke'aximum of four hours during reqrrired operability testing of the HPCI system, RCIC system and the dryvell-pressure suppression chamber vacuum breakers.QCYIohK 6+8 b.If the differential pressure of Specification 3.7.A.6.a cannot be maintained arid the differential pressure cannot be restored vithin the subsequent eriod Fcd~rNcM re4L po~gg (5%ii l2 I<rS BFH Unit 2 3.7/4.7-12

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Al ambe a.Differential pressure betveen the dryvell and suppression chamber shall be maintained at equal to or greater than 1.1 psid except as specified in (1)and (2)belov: a.The pressure differential betveen the dryvell and suppression chamber shall be recorded at ast once eac sh~82 ls hoclr$(1)This differential shall be established vithin 24 hours of a ev~fIi%b;l,'W) erat ng temperature and pressure The erent a pressure may be reduced to less than 1.1 si 24 hour r or to a scheduled utdovn.LI Leo Ao&I (2)is differential may be decreased to less than 1.1 psid for a maximum of four hours during required operability testing of the HPCI system, RCIC system and the dryvell-pressure suppression hamber vacuum breakers.ACrloq b.If the differential pressure of Specification 3.7.A.6.a cannot be maintained and the differential pressure cannot be restored vithin th subsequent ska-hour pQri&h~(.=Rc~lllcc~1m~poult'R Q~(y+J~Ih JQ.h~i s, BFH Unit 3 3.7/4.7-12 P<QE o JUSTIFICATION FOR CHANGES BFN ISTS 3.6.2.6 DRYWELL-TO-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE ADMINISTRATIVE CHANGES Al A2 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The Frequency for verifying the pressure differential between the drywell and the suppression chamber has been changed to 12 hours from shiftly.CTS Table 1.1 defines shiftly as at least once per 12 hours.As such, this is a change in presentation only and is therefore administrative. TECHNICAL CHANGES-LESS RESTRICTIVE Ll L2 The proposed change revises the required initiation point for establishing differential pressure between the drywell and suppression chamber.By increasing the initiation point following startup to 15%rated thermal power (RTP)(CTS initiation point is operating temperature and pressure, which is about 1%RTP), the drywell pressure and temperature will have sufficient time to stabilize prior to establishing the required differential pressure.As long as reactor power is below 155 RTP, the probability of an event that generates excessive loads on primary containment occurring within the first 24 hours of a startup or within the last 24 hours before shutdown is low.24 hours is considered a reasonable amount of time to allow plant personnel to establish the required differential pressure.CTS 3.7.A.6.b allows 6 hours to restore the differential pressure before initiating an orderly shutdown, which requires the plant to be in Cold Shutdown within 24 hours.The proposed actions allow 8 hours to restore differential pressure and 12 hours to reduce thermal power to<151 RTP.Below this power level, per the proposed Specification, the LCO is no longer applicable (See Comment Ll above).PAG~~OF~BFN-UNITS I, 2, 5 3 Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP

(g'i APA 2 9 1993 3.7.F.est CR 3.7.F.3 (Continued) these primary containment isolation valves is governed by Technical Specification 3.7.D.ScŽ3 L5Hfs cR hoA Qr Changers Ae cT5 p q p/q p~b.Prcssure control of thc containment is normally performed by VEHTIHQ through 2-inch primary containment isolation valves vhich route effluent to the Standby Gas Treatment System.The OPERABILITY of these primary containment isolation valves is governed by Technical Specification 3.7.D.Qi 1.The Containmcnt Atmosphere Dilution (CAD)System shall bc OPERABLE with!a.Two independent systems capable of supplying nitrogen to the dryvell and torus e SR zc.~,t.(b.A miniam supply of 2,500 gallons of liquid nitrogen pcr systems 5 R 3,v,z.t.2 a.clc ach s len id 0 crate air t gcn v ve ough t le t on comp te cyc of ull t vcl in a cord cc vit ec fication 1.0.MK and at least once per month verify that each manual valve in the flow path is open.b.Verify that the CAD System contains a minimum supply of 2,500 gallons of liquid nitrog vic-er week.BHf Unit 1 3.7/4.7-22 mMmrN.Zsg PAGE~OF~ 2 ccig+for)g, Q, p J DEC 07 1994 2.~z.c.Z.I+Applt cab;la4)The Containment Atmosphere Dilution (CAD)System shall be OPERABLE whenever the reactor is in the RUN MODE or g~~+a/M 2.2.When FCV 84-8B is inoper-able, each solen id operat d air/nitr en valve o System B all be cycle through at least one piete cycle f full trave and each m ual valve in the flow path of System B shall be verif open at least once per week.ifc7/o 4 3.If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE.~~<M r<d/f2.Bean A,(MT(od 8 4.If Specifications 3.7.G.1 and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in Pln pE 3 l4/Olin/2 I}cars 5.Pr ry co tainme pre ure shal be 1 ted to axim of 30 sig d ing r ress izati folio ng a los of c olant cident.LA 2 6.System may e cons ered OPERABLE with FCV 84-8B inoperable prov ded that all active component in Syst B and all o er'ctive mponents in System A e OPERABLE.7.ecification 3.7.G.6 and 4..G.2 are in feet until the rst Cold Sh tdown of unit 1 fter July 20, 1984 or until January 17, 1985 whichever occurs first.BFN Unit 1 3.7/4.7-23

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

5 ecjgiea4ion 9 6.3: I APR 2 9 t991 3.7.F.a Co a~St g 3.7.F.3'(Continued) 4.7.F.~Sst g ot et ue 0 these primary containment isolation valves is governed by Technical Specification 3.7.D.b.Pressure control of the containment is normally performed by VENTING through 2-inch primary containment isolation valves vhich route effluent to the Standby Gas Treatment System.The OPERABILITY of these primary containment isolation valves is governed by Technical Specification 3.7.D.os e e P,)Qt,'0 3.6,3.l 1.The Containment Atmosphere Dilution (CAD)System shall be OPERABLE vith: 4~~~Sec~~s4;4';(~4; 4r e4)e 4 r c.YS S.v.F/47,F CAD a.Tvo independent systems capable of supplying nitrogen to the dryvell and torus.~~p,&,S.l.(b.A minimum supply of 2,500 gallons of liquid nitrogen per system.a.Cycle e ch solenoid crated air/n rogen va ve thr gh a lea one c piet cycle f full ravel in acco dance vith S ecif n 1.0.MM A/3.4.~.<+and at least once per month verify that each manual valve in the flov path is open.gg,g, l I b.Verify that the CAD System contains a mini&urn supply of 2,500 gallons of liquid nitro e vice per vee BFN Unit 2 3.7/4.7-22 AMENDMENT go.y 9 7 pAGE GF

5 cci ic44I 3.cn.3.I 0)g 07 1994 Al LCo 3.4Y.l 2 Appfchl,4$ $P Cl7oN A The Containment Atmosphere Dilution (CAD)System shall be OPERABLE whenever the reactor i.s in the RUN MOD er SgAN'Tup HoD6 HZ If one system is inoperable, the reactor may remain in*operation for a period of 30 days provided all active components in the other system are OPERABLE.Q~Acko~A 4 Ac.i (os If Specifi.cations 3.7.G.1 and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in]claps 3 i~5.Primary c tainment pressure 11 be 1 ted to a um of 30 ig durin repr surization llowing loss o coolant accident.BFN Unit 2 3.7/4.7-23 NmwBtr HO.229 pAcs 0F

UNIT 3 CURRENT.TECHNICAL SPECIFICATION MARKUP SEMI elk'hD 3 4 3 J 4'9)9g)3.7.F.4.7.F.3.7.F.3 (Continued) these primary containment isolation valves is governed by Technical Specification 3.7.D.SCr 3<g+Cl aa$o n Q Qz+~<~C T$3,7.F'/9 I F b.Pressure control of the containment ls normally performed by VERTIRQ through 2-inch primary containment isolation valves vhich route effluent to the Standby Gas Treatment System.The OPERhBILITY of these primary containment isolation valves is governed by Technical Specification 3.7.D, 1.The Containment Atmosphere Dilution (CAD)System shall be OPERABLE vith: a.Two iILdependent systems capable of supplying nitrogen to the dryve11 and torus o ae cle ach lenoid crate air trogen, v ve ough a le t one comple e cycl of f l,tra el in accordance vlth o.MM BPK unit 3 b.k minimum supply of 2,S00 gallons of liquid nitrogen per systeme and at least once per month verify that each manual valve in the flov path is open.54 3.4P'.l b.Verify that the CAD System contains a minhmun supply of 2,500 gallons of ll id nitrogen ici per ve 3.7/4.7-22 'IIBfllfgff NO, g g g PAGE~OF~ 0 5 o.Ai EC 0 7 1994 2.Lco 7,4.3,t]g4o q The Containment Atmosphere Dilution (CAD)System shall be OPERABLE whenever the reactor is in the RUN II 2 o~A'soup wove If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE.Ilotc A~$4stwd hcVion A,l RZ 4 lkho w 8 If Specifications 3.7.G.l and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in N/ID'NI fht g/Q~l 5.Prima cont'nment p essuxe hall b limit d to a imum 30 p j dur g rep essuri tion ollowi a loss f cool t ac dent.Lgz BFN Unit 3 3.7/4.7-23 AMENDMENT NO.I 8 6 phd E 3 t JUSTIFICATION FOR CHANGES BFN ISTS 3.6.3.1-CONTAINMENT AIR DILUTION SYSTEH ADMINISTRATIVE CHANGES A1 A2~A3 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A NOTE was added specifying LCO 3.0.4 is not applicable. Since the current Technical Specifications do not have LCO 3.0.4, stating it is not applicable constitutes an administrative change.Unit 1 CTS 3.7.G.6 5 7 and 4.7.G.2 have been deleted.These Specifications were special provisions that expired January 17, 1985, and therefore, no longer apply.As such, the proposed deletion is considered administrative. TECHNICAL CHANGES-MORE RESTRICTIVE The Surveillance Requirement has been revised to include each manual, power operated, and automatic valve that is not locked, sealed, or otherwise secured in position.H2 This change adds MODE 2 (STARTUP NODE)to the Applicability to go along with MODE 1 (RUN MODE)which is already required.The CAD System is required to maintain the oxygen concentration in the primary containment below the flammability limit following a LOCA.Adding a new MODE to the Applicability constitutes a more restrictive change.This change is consistent with NUREG-1433. BFN-UNITS 1, 2, 8L 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.3.1-CONTAINMENT AIR DILUTION SYSTEM Proposed ACTION C is more restrictive since it requires the unit to placed in MODE 3 in 12 hours versus CTS 3.7.G.5 which requires that an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION with'in 24 hours.In addition, since the existing Specification (CTS 3.7.G.2)is only applicable during the RUN mode (MODE 1), failure to meet the existing specification would only require the unit be placed in at least STARTUP/HOT STANDBY (MODE 2)in 24 hours since at that time CTS 3.7.G.2 is again met.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1~LA2 This Surveillance is being relocated to plant procedures (IST program)since these valves are tested as part of the IST program.As such, it is not needed to be specified as a specific Surveillance Requirement. If during testing or routine use of the system they are found to be inoperable, the appropriate ACTIONS would be taken.This change is consistent with the BWR Standard Technical Specifications, NUREG 1433.This requirement has been relocated to plant procedures. This type of action is a post-accident action routinely governed by the emergency operating procedures. Any changes to the procedures would be controlled by the licensee controlled programs."Specific" I Ll The Frequency of this Surveillance has been extended to 31 days, similar to other surveillances on tank content (e.g., diesel fuel oil).The nitrogen tank contents only decrease when nitrogen is being added to the drywell, and this evolution is a manually actuated and secured evolution (i.e., it is a very controlled evolution). If nitrogen was being added, it would be monitored more closely.Thus, since there are very positive means to ensure nitrogen tank volume is monitored if being used, and volume does not decrease due to"automatic, unmonitored" use, the 31 day Frequency is considered appropriate. BFN-UNITS 1, 2,&3 Revision 0 CH 3.7.4.4 (Conc, 4.7.L.4 (Cans'd)c.zo dryveLL-suppression chamber vacuum breakers may be determined to be oyerable fox opening.c."=ach vac~breaks valve shall be insyecccd for propex ayeracian of the valve lnd lent sW-ches in accordance vith Spec'f'cat,'oa L.Q..C.See Yuu,+h>>I'~F-8~<Ts 3417 4.If Spec'f'cac'ans 3.7.4.4.a, 3.7.4.4.b, or 3.7.4.4.c. cannoc be met, the unit shall be placed ia a COLD SHUTDOWE COND~OH in aa order'y mana 24 hours.d.4 leak test of the dz:~el ta suyyressian chamber structure shall be conduct 4 during each oyerasiag cycLe.Acceptable leak rate is 0.09 lb/sec eC'rimary consainmeat, atmosphere vich si dif~a tacial.Lfo~A3.a.Cantainmeac acmosyhere shall be reduced to less thaa 4Z oxygen v as 4uring reaccox yovex o erati re+t oa x'LOAN ig>>except as specified in 3~eke~~Se 3.C.a.z, t a.The yrimary containmenc oxygea coaceatration shall be measured c il oxyg ea s 1 be ad]ced so acc unt or~SI'in of by ad ing redetersdn in b.Ql'eh'.lip Vi~the 24-hour yeriod subsequent co lacing the reactor in the.D folloving a shut dova, the containment atmosphere oxygen concentrasioa shall be reduceci ro less than 4 by volume maintained in this condition. b.The tho us to asure~p imary cant inmen o gea conc erat, an'l be cali ased nce very efueling cycle Deiaex'cing may commence prior to a shutdova.ur L43 C~.fy ro a r is be used suyyl the pa cic caa rol s tea'e pr caacai eac, th reacco shall n s be sca ed, ar ac po r, the actor sh 11 be b ught to a COLD OMÃCOHDIT ON vf.~a 24 hours.C.The contxol air supyly valve or the pneumat c con~o1 cern inside ch primary ca ainment, shall e ver.'fied clos priox to rea or star=.p 4 monthly the eaft d.1 Acrtod 5 BFH Unit 1 If Specificatioa 3.7.h.5.a d 3.7.A.5.b cannot bc mec su ovasa be initiated aad the reaccor shall be in a COLD SHUTDOWN CONDITION vithia 24 hours 3.7/4.7-11 P(op W T]o~A L2 AMENOMENT>to y>9 PAGE~OF~. NOV 22 t888 S ec,4~6<~a~,Z S."-OR 5P RPiIOH 3.7.A.4 (Coat'd)4 7.L 4 (Cont'd)c.Tvo dryve3~-suppression chamber vacuum breakers may be determined to be inoy'erable for oy~Co Each vacuum breaker, valve shall be insyectcd'or proper operation of."c valve aacL Limit svit"'"es in accordance vith Specif icatioa 1.0.%$.5'8Z ZVs7IPICgg~g FOR CPANCrES FOR BFN O'TS Z.g.l.7 d.Zf Syecificatioas 3.7.4.4.a,.b, or.c cacaos be met, the unit shall be ylaced in a Cold Shutdovn condition ia aa orderly manner vithia 24 hours.k leak test of the d~ell to suyyressioa chamber structure shall be conducted duri1sg each operating cycle acceptable leak rate L's 0.09 lb/sec of primary, containmcat atmosphere vith 1 psi differential. L Cg 3.$.53-a.b.A~(ie<L;l,<g Containment atmosphere shall be reduced to less than 4X oxygen v ro as duriag reactor over oyeratio pith~eacror c 1 t sacs'lboveMOO~s except as specified in 3.7.i..b.Ll Mithin the 24-hour pcr subsequent to y aciag the reactor in the UH mo folloviag a shut-dova, the coatainment atmosyhere oxygen conccatration shall be reduced to less than 4X by volume aad maintained ia this condition. einert may commence 4 hour yrior to a shutdovn.measurem shall b adjusted t ccount'he unc tainty of methocL ed by ad g a predetermined error funct.'o The thods ed to casus thc y ry c tainme gen acentr ion sh 11 be alibr tcd onc every refuel S~~>>~'a.The primary'ontainment oxygen concentration shabe measured d c~dail e oxygen L42 LS c.Zf plant control air is being used t supply the tic coa ol sys em inside pr cont t, the actor shall a t be start or if pover, the eactor shall be brou t to a Cold utdovn coaditioa vithin 24 hours.co The control air suyyly valve f the pneumatic control sys cm inside thQ prima coat cat shallibe ver ied close rior to reagtor s rtuy and mon y thereafter. 67/54 8 d.If Specification 3.7.JL..a and 3.7.'A.5.b canaot be me order y s ut ovn s e initiated and the reactor shall be in a Cold Shutdovn conditioa vithia.24 hours.a~An-(oN A BFH Unit 2 3.7/4.7-11 Ah!B~DMB"~~~.yg~PAGE J~<t NOV 22 Icy;3.7.A.4 (Cont'd)(4.7.A.4 (Cont'cc.Qu5+gc4gA) Oa~Pa QFt0(5g>g.g~,g c.Two drywcll-suppression chamber vacuum breakers may bc determined to bc inoperable for opening.d.If Specifications 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c, cannot be met, the unit shall bc placed in a Cold Shutdown condition in an orderly manner within 24 hours.c.Once each operating cycle, tach vacuum breaker valve shall bc inspected for proper operation oi the valve and limit switches in accordance with Specification 1.0.MM, d.A leak test of thc drywcll to suppression chamber structure shall be conducted during each operating cycle.Acceptable leak rate is 0.09 lb/scc of primary containment atmosphere with 1 si differential. I Lac 3,e.v.z a.Containmcnt atmosphere shall be reduced to less than 4X oxygen t o as during reactor power operation t a~o res ure ove 00+sig except as spec e n 3.7.A..b.5g 3 b,X 2.l a.The primary containment oxygen concentration shall L7 be measu o d WY dail e oxy en e urem t sha be a ust d to a coun for th unce ain y~Z of th meth d used ad g a pred rmi d error function b..Within the 2 our period subsequent to placing the reactor in the UH mod following a shut-~down, thc containment atmosphere oxygen concentration shall be reduced to less than 4X by volume and in this condition. 4 Deinerting may commence 24 hours to c meth s used to me urc th prima coats cnt oxy n conc ntrati shal be ca ibrate once e ry refuel cycl.C~plan con ro a s e ng used o supply e pneuma c control s tem insi primary ntainment, the reactor s ll not be tartcd, or i at power, he rcacto shall be bro t to a C d Shutdo condition within 24 ours.c.control ai supply valve fo thc pneumati control syst inside thc imary contai ent shall be rifi closed p r to reactor startup d monthly thercaftcr. d.If thc specifications of 3.7.A..a through 3.7.A.5.b cannot bc met Plop st gh'Ogg l2 khon B BFH Unit 3 an or cr y s ut own shall bc initiated and the reactor shall be in a Cold Shutdown condition within 24 hours.3.7/4.7-11 PAGE~OF~AMENDV:E~lT N~.18 C' AlUSTIFICATION FOR CHANGES BFN ISTS 3.6.3.2 PRIMARY CONTAINMENT OXYGEN CONCENTRATION ADMINISTRATIVE CHANGES Al A2 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.This statement has been deleted since it is unnecessary. With the reactor in power operation, reactor coolant pressure will always be above 100 psig.TECHNICAL CHANGES-MORE RESTRICTIVE The requirement to place the plant in Cold Shutdown condition within 24 hours when the limit is not restored within the required Completion Time is revised to reflect placing the plant in a non-applicable condition. CTS 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in Cold Shutdown is not applicable if thermal power is reduced to<15%RTP (outside the applicable condition) within 8 hours.The current action allows 24 hours to place the plant in a non-applicable condition. As such, this is an additional restriction on plant operation. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl~LA2 The details of how to reduce oxygen concentration to less than 4%have been eliminated from the ISTS.This type of detail will be retained in plant procedures and/or system operating instructions. Details on the methods of performing sorveillances has been relocated to BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF JUSTIFICATION FOR CHANGES BFN ISTS 3.6.3.2 PRIMARY CONTAINNENT OXYGEN CONCENTRATION LA3 LA4 plant procedures. Changes to plant procedures will be controlled by the licensee controlled programs.Requirements for controlling the use of plant control air to supply the pneumatic control system inside the primary containment and the associated surveillance have been relocated to the Technical Requirements Manual (TRN).The requirement to record the containment oxygen concentration will be relocated to plant procedures. Changes to plant procedures will be controlled by the licensee controlled programs."Specific" Ll L2 The 24 hour allowance for inerting on startup has been changed to allow 24 hours after exceeding 15%power instead of the current Run Mode requirement (approximately 5%).The 24 hour allowance for de-inerting on shutdown has been changed to allow 24 hours prior to reducing below 15%power.These small differences provide some added time to inert or de-inert the drywell, and provide consistency with BWR Standard Technical Specifications, NUREG-1433. These minor changes are justified, since the time allowed without an inerted drywell is.only increased slightly, and the fact that at low power levels, hydrogen generation is very small compared to higher power levels.Currently, no time is provided to restore oxygen concentration to within limit prior to requiring a plant shutdown.Proposed Required Action A.1 and associated Completion Time will allow 24 hours to restore oxygen to within the limit prior to requiring a plant shutdown.During this time, the CAD System is normally still OPERABLE, thus a means to prevent combustible mixtures still exists.This new ACTION would possibly prevent unnecessary shutdown and the increased potential for transients associated with the shutdown.L3 The periodic verification of oxygen concentration in the primary containment has been changed from a daily verification to a weekly verification. The primary containment is inerted to maintain oxygen concentrations within limits.The primary containment leak rate is established for each operating cycle and any changes during normal operation usually occur very slowly.Other changes to primary containment integrity, such as PCIV operability problems, are indicated by other means to the plant operator and appropriate actions are contained in other technical specifications. BFN-UNITS 1, 2, 5 3 Revision 0 PAUE~OF~ UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 4 4.7.B.3.7.B.4 (Cont'd)b.Place all reactars in at least a HOT SHUTDOWÃCOHDITIOS vithin thc next 12 hours and in a COLD SHUTDOWÃCOIITIOH vithin the follovi 24 hours.sc<Duswg'~So n*r changes A e~~iSrs Z.s,q.p reactor zone at all except as specified 3.7.C 2~times in>~<81&~&lTlnN g+~Ofba bl~Secondary containment W~~g, 4RmSRiP 1.Secondary containment surveillance shall be performed as indicated belov: ,gq, 5 3b9('k a.Secondary conta nmcnt capability to maintain ,1/4 inch of vater vacuum~l'ca vi i uw'~h)on 4ns P2 vith c ra of not ore tha 12,000 cfm sh dcmonstratc t cych r ue ng ut e Qio r Lhmw 2.Af r a s condary ca tai nt v lation is~2 term ed, he sta dby gas teat nt s stem v ll bc oper cd i ediat ly a ter the ffec ed zon s ar iso ated from t e re ainde of the econda c tai cnt t con irm i abili to m inta n th rema ndcr o the scco ary can inmen at 1 4-i h of vater egati e pr ssurc under calm vind con itions 2.If reactor zone secondary containment i,aeeg+44y<cannot be maintained the folloving conditions shall be met: N~k.4 ecu;t'., I a.Suspend all fuel handling operations, core altera-tions, and activities vith the potential to drain any reactor vessel containing fue.t~aebwk(Lz.A-CT(on/C AMENDMENT NO.I 74 3.7/4.7-16 BFS Unit 1 b.Restore reactor zone secondary containmcnt AC7(g~5 integrity vithin 4 hours, or place all rcsctars in at least a HOT SHUTDOWN COHDITI05 vithin the next 12 hours and in a COLD SHUT-DOWN COHDITI01 vithin the folloving 24 hours. Cl 3 Secoada containment integ-rity s 11 be main ined in the r ueliag xone except as spec ied in 3.7..4.4.IS refueliag x c secon ry taiament c ot be atained t followi g conditions ll be t: a.Handl g of s t fuel and all rations over spent fue pools an open re ctor wcl s contai ing fuel shall be prohibit d.b.The stan y gas tr atmeat system ction to he refuel g xone w 1 be block d except r a cont oiled lea c are six to ass the ach evtag of vacuum of at least 1/inch of water and aot ov r 3 inch s of water ia 1 three reactor xones.is is y appli-cable ii reactor one integrity is required.&e xush'f~~~A~C~Q~SF'SrS, 3.4,t.p D.D 1.Rhea Primary Containmeat. Integrity is required, all primary containmeat isolation valves and all reactor , coolant system instrument line flow check valves shall bc OPERhBLEc except as~pecified ia 3.7.D.2.+Locked or sealed closed valves may be opened on an inter-mittent basis under istrative control l.The primary containment isolation valves surveillance shall bc performed as follows: a.ht least once pcr oper-ating cycle, the OPER-.hBLE primary contain-ment isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiati bFN Unit 1 3.7/4.7-17 AMENMER NL I 8 9 PAGE OF I~ UNIT 2 CURRENT.TECHNICAL SPECIFICATION MARKUP Cl 0 S i4co4iam 3.l.Q.t MAR 30 1%0 3.7.B.S t 4.7.B.ta db Gas ea$2U: tt9),3.7.8.4 (Cont'd)b.Place all reactors in at least a HOT SHUTDOMH COHDITIOH vithin the next 12 hours and in a COLD SHDTDOWH COHDITIOH vi thin the folloving 24 hours.3.7.C.S o d Ca a e t 1.Secondary containmen shall be in the reactor zone at all times except as specified in 3.7.C.2.oPenRB~govwn<oeJ A+~Qc,Suf,4i(i cak>e~for C~~~pg Qr9t td iSTS R,C, k3 4.7.C.Seep da Co ta nt Ll 1.Secondary containment surveillance shall be pcrfarmed as indicated bclov:>>q,(,3+sR3 econdary containment capability to maintain/p(1/4 inch of water vacuum er c~v ()L'i~~(<m h bandit on i%a~HR v th a system leakag rate of not more tha 12,000 cfm shall be demonstrated ze uagngou~ge >0)houP-LAO to rc cli 2.If reactor conc secandary containment 4ae~~cannot be maintained the folloving Lo-c ns shall bc met: Re i~Ae4ia a.Suspend all fuel handling QL<(ot4 apcratians, core altcra<<t'ions, and activities vith the potential to drain any reactor vessel containing i~ka4Q b.Restore reactor zone A~<" secondary cantainment A<@integrity vithin 4 hours, or place all reactors in at least a HOT SHUTDOWH COHDITION within the next 12 hours and in a COLD SHUT-DOW COHDITIOH vithin the folloving 24 hours.2~Aft a se ondary co tainme t viola on is termi d, the tandby gas treat nt sys m vill b oper ted i diately fter, th affec d zones re i olated from th remainder f thc econda conta cnt t confirm ts abil ty to intain e remainder f the se anda containm t at 1/inch of wate negativ pre ure under calm vind conditions. 4 SPY 3.4-l l a,~dL z.c, V.(.BFH Unit 2 3.7/4.7-16 AMENOMENT gp.y 7 7 PAGE~OF~

~p<<Aicc~io~ 3.6.4 (SSES 3.Secondary c ntainment integ-rity shall e maintained in the rcfu ing zone, except as, specifi d in 3.7.C.4.4.If r ueling zone secondary con ainment cannot be m ntained the folloving nditions shall be met.a.Handling of spen fuel and all operations ver spent fuel pools open reactor veils conta ing fuel shall bc prohib ed.b.The st dby gas treatm nt syst suction to the ref cling zone vill e b eked except for a ontrolled leaks area sized to assure the achieving, of vacuum of at least 1/4 inch of vater and not ov 3 inches of vater in ll three reactor zones.is is only appli-cable if reactor zone integrity is required.Sec, guS g Cr+A'on Crt A>>gCgg*~Ben/isrS Z.C.l.g a Co ta a o D.ma Co ta e t Isolat o V~a~1.Shen Primary Containmcnt Integrity is required, all primary containment isolation valves and all reactor coolant system instrument line flov check valves shall be OPERABLE*except as specified in 3.7.D.2.*Locked or sealed closed valves may be opened on an intermittent basis under administrative control.1.The primary containment isolation valves surveillance shall be performed as follovs: a.At least once per oper-ating cycle, the OPER-ABLE primary contain-ment isolation valves that are povcr operated and automatically initiated shall be tested for simulated automatic initiation BiH Unit 2 3.7/4.7-17 NENOMEHT NO.2 04 CURRENT TECHNICAL SPECIFICATION MARKUP , 3.7.B.S s.B.a kama eat ent 3.7.B.4 (Cont'd)b.Place all reactors in at least a HOT SHUTDOWH COHDITIOH within the next 12 hours and in a COLD SHUTDOWH COHDITIOH vi thin hc folloving 24 hours.S<~wustj$i~*'o r)CholcS Fo r GV~ISTs 3oL Qe3 CoaD'Ihonl g+c fcrobl8 3.7.C.Seco da Co ta e shall be ma nta nc n the reactor zone at all imes>Pal'caYiliFy 'xcept as spccificd in 3.7.C.2.PcRAbl CO not app ca e prior to load ng fuel into c Unit 3 rca tor vcs 1, pr vided the Uni 3 rcac r zone s not requir for second containment integrity for other 4.7.C.Seep da Containment 1.Secondary containment surveillance shall be performed as indicated below: SR 3.4.4I.3 8-sR.o.l.a.Secondary containment capability to maintain 1/4 inch of vater vacuum er aim vn lOi~n mp c di on vith system nlea ag rate of not more tha 12,000 cfm shal emonstrate at eac e e ng o tage rior o fu i RC&n>R.+B b.Restore reactor zone secondary containment integrity vithin 4 hours, or place all reactors in at least a HOT SHUTDOWH COHDITIOH vithin the next 12 hours and in a COLD SHUT-DOWH COHDITIOH vithin the folloving 24 hours.2.If reactor zone secon ary containment iaeegekty cannot bc maintained the folloving ons shall bc mct: r+pesek Note 4o n C.)a.Suspend all fuel handling operations, core altera-tionsp and activities with thc potential to drain any reactor vcsscl containing fuels tA&il Fel After scco dary cont nmen viol tion is det rminc , the stand gas tr atm sys vil be o cra d imm iatel after the fecte zones are iso atcd f om the remainder of hc sc ondary contai nt to onfi its abilit to ma tain e'emai der of he se onda cont inmcn at 1/inch f vater egativ pres re dcr ca vind onditions/tdd Sls Xg,q,~,~~3i6,$, t,+BFH Unit 3 3.7/4.7-16 NENOMENT NO.1 5 9 WGE~OF~

sPec,,g z.s.M.t econdary con ament integ-rity ahal e maintained in the re ling xone, exec as spec led in 3.7.C.4.5 refueling xon secondary containment c ot bc maintained e folloviag conditio shall bc met: ae ling of sp fuel and all operatio over spent fuel pools open reactor veils co aining fuel hall bc pro bited.b.Th standby g treatment stem suet n to thc refucll one vill be blocke except for a contr led leakage are six d to assure the a eving, of a va of t least 1/4-in of vatcr and not over inches of vater in a three reactor xoncs.s is only appli-cable i reactor zone intcgr ty is required.$c'w Fu~f4 cdjog Qr Cjg~W 8%lsT5 3.t.I.3 D D.spa o 1.When Primary Containment Integrity is required, all primary containment isolation valves and all reactor coolant system instrment line flov check valves shall be OPERABLE*except as specified in 3.7.D.2.<<Locked or sealed closed valves may bc opened on aa latczmittent basis under admiaistratlve control.1.Thc primary coataiameat isolatioa valves surveillance shall be performed as follovs: a.At least once per oper-ating cycle, the OPER-ABLE primary contain-ment isolation valves that are povcr operated and automatically initiated shall be tested for simulated automatic initiation BPK Unit 3 3.7/4.7-17 AMENDMENT,Ng g 6 I

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.1-SECONDARY CONTAINMENT ADMINISTRATIVE CHANGES Al A2 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The definition of SECONDARY CONTAINMENT INTEGRITY has been deleted from the proposed Technical Specifications. In its place the requirement for secondary containment is that it"shall be OPERABLE." This was done because of the confusion associated with these definitions compared to its use in the respective LCO.The change is editorial in that all the requirements are specifically addressed in the proposed LCO for the secondary containment and in the Secondary Containment Isolation Valves and Standby Gas Treatment System Specifications. The Applicability has been reworded to be consistent with the new definitions of NODES and to have a positive statement as to when it is applicable, not when it is not applicable. Therefore the change is purely a presentation preference adopted by the BWR Standard Technical Specifications, NUREG-1433.A3 Amendment 159 to Unit 3 Technical Specifications added a provision to allow separating the Unit 3 reactor zone from the secondary containment envelope under certain conditions (prior to fuel loading)to expedite Unit 3 constructions activities'during Unit 2 operation. This provision is no longer needed and can no longer be applied.Therefore the*Note to TS 3.7.C.1 has been deleted.This change is considered administrative since it deletes a requirement that no longer applies.TECHNICAL CHANGES-MORE RESTRICTIVE Ml This Surveillance (it appears to be only one Surveillance, though it is in two parts)has been broken into two separate Surveillances, SR BFN-UNITS 1, 2, 5 3 Revision 0 PAGE~OF~

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.1-SECONDARY CONTAINMENT H2 M4 3.6.1.4.3 and SR 3.6.1.4.4. The tests will ensure the ability of the secondary containment to maintain I/4 inch vacuum, and in addition, SR 3.6.4.1.3 will ensure the vacuum is attained in 120 seconds, while SR 3.6.4.1.4 will ensure it maintains the vacuum for I hour.These new requirements are additional restrictions on plant operation. The analysis for secondary containment drawdown assumes two SGT subsystems are needed.Thus, the test now specifies the minimum number of'operating SGT subsystems and the total flow rate.To ensure all three SGT subsystems are tested (since the test does not specify that all SGT subsystems must be tested)the Frequency is on a STAGGERED TEST BASIS, which will ensure all three SGT subsystems are tested in 2 cycles.These are additional restrictions of plant operation. Two new Surveillance Requirements have been added.SR 3.6.4.1.1 will verify that all secondary containment hatches are closed and sealed every 31 days.SR 3.6.4.1.2 will verify that each access door is closed, except when used for opening, and then one door is closed, every 31 days.These are additional restrictions on plant operation. This change requires the movement of irradiated fuel in secondary containment and CORE ALTERATIONS to be"Immediately" suspended if secondary containment is inoperable. In addition,,action must be"Immediately" initiated to suspend operations with the potential to drain the reactor vessel in this Condition. The current specification does not establish a time limit to suspend these activities. Immediately suspending these activities minimizes the probability of a fission product release if a reactivity event occurs while the secondary containment is inoperable. Also, immediately initiating action to suspend operation with the potential to drain the reactor vessel will minimize the potential for reactor vessel draindown and subsequent potential for fission release.Imposing a time limit to suspended these activities is a more restrictive change.The reactor building is divided into four ventilation zones which may be isolated independently of each other.The refueling room which is common to all three units forms the refueling zone.The individual units below the refueling floor form the other three reactor zones.The zone system is not an engineered safeguard, and the failure of the zone system would not in any way prevent isolation or reduce the capacity of the Secondary Containment System.If the internal zone boundaries should fail, the entire reactor building still meets the requirements of secondary containment. CTS 3.7.C requires, the secondary containment integrity to be maintained in the reactor zone and refueling zone at all times except as specified in 3.7.C.2 and 3.7.C.4 respectively. If secondary containment cannot be maintained in the reactor zone, fuel BFN-UNITS I, 2, 5 3 Revision 0 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.1-SECONDARY CONTAINMENT secondary containment must be restored within 4 hours or all reactor shall be shut down.If secondary containment cannot be maintained in the refueling zone, the handling of spent fuel and all operations over spent fuel pools and open reactor wells shall be prohibited. Currently, a combined secondary containment integrity test is performed to demonstrate Technical Specification operability. In addition, due to leakage between zones, zone integrity is difficult to maintain.As such, secondary containment integrity is maintained on the three reactor zones and the refueling zone at all time.Therefore, the separate Specification that only prohibits the handling of spent fuel and all operations over spent fuel pools and open reactor wells when refueling zone integrity is not maintained is not necessary and has been deleted TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1 This design detail/requirement has been relocated to the Background section of the Bases for ITS 3.6.4.3,"Standby Gas Treatment System," and to plant procedures governing this Surveillance Requirement. Any changes to this requirement will require a licensee controlled program evaluation. LA2 The requirement to operate the Standby Gas Treatment System after a secondary containment violation is determined and has been isolated (i.e., restored)to check if it can maintain the proper vacuum is being relocated to plant procedures. Any time the OPERABILITY of a system or component has been affected by maintenance, replacement, or repair, post maintenance testing is required to demonstrate OPERABILITY of the system or components. Explicit post maintenance surveillance testing has therefore been deleted from the Technical Specifications and will be relocated to the appropriate plant procedures. Any changes to the requirement will require a licensee controlled program evaluation. This change is consistent with NUREG-1433."Specific" L1 The proposed surveillances for the 1/4 inch vacuum tests do not include the restriction on plant conditions that requires the surveillances to be performed during a refueling outage, prior to refueling. These Surveillances could be adequately performed in other than a refueling outage without jeopardizing safe plant operations. The control of the plant conditions appropriate to perform the test is an issue for procedures and scheduling, and has been determined by the NRC Staff to BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION fOR CHANGES BFN ISTS 3.6.4.1-SECONDARY CONTAINMENT plant conditions appropriate to perform the test is an issue for procedures and scheduling, and has been determined by the NRC Staff to be unnecessary as a Technical Specification restriction. As indicated in Generic Letter 91-04, allowing this control is consistent with the vast majority of other Technical Specification surveillances that do not dictate plant conditions for the surveillances. The proposed change to the 18 month frequency also effectively increases the surveillance interval.The current Technical Specification for all three units requires performance at each refueling outage prior to refueling. Since the secondary containment is common to all three BFN units, with all three units operating, this could result in performance of the same test at an average of every 6 months.The change to the 18 month frequency will allow this test to be performed once and applied to all three units Technical Specifications. Since operating experience has shown these component usually pass the Surveillance at the 18 month frequency, the frequency is considered acceptable from a reliability standpoint. L2 Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operation and the inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.By adding an exception to LCO 3.0.3 for the failing to suspend irradiated fuel movement, an LCO 3.0.3 required reactor shutdown is avoided in MODE 1, 2, or 3.However, the plant would still be required to shutdown after 4 hours per proposed Required Actions B.l and B.2 in addition to suspending fuel movement per Required Action C.l.BFN-UNITS 1, 2, 5, 3 Revision 0 PAGp~OF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION It MARKUP PAGF~OF~

~3 6.Y-2.MA 80 1980 SURVBILLASCE REQUIRENESTS 3.7.B.4.7.B.3.7.B.4 (Cont'd)b.Place all reactors in at least a HOT SHUTDOMH COIITIOK vithln the next 12 hours and in a COLD SHUTDOMf COlIDITI01 vi thin the folloving 24 hours.SC'34s~A cw6<hr C~cs@~BFrd t 5 TS Z, C./.Z 4.7.C.f~seat L4p 3,&Ac%+~i:cab'3i Secondary contalnmcnt integrity shall be maintained in the reactor zone at all times except as specified in 3.7.C 2'jiopogCg(hfo~P+3 k+H)Oq5 AF f~.~~~nkk l h+6o~s 1.Secondary containment surveillance shall bc performed as indicated bclov!a.Secondary containment capability to maintain 1/4 inch of vater vacuum under calm vlnd (<5 mph)conditions vlth a system leakage rate of not more than 12,000 cfm, shall be demonstrated at each refueling outage prior to refueling. If reactor zone secondary, containment integrity cannot be maintained the folloving conditions shall be mct: 2~4 kz~~X~g p.a.Suspend all fuel handling operations, core altera-tions, and activities vith the potential to drain any reactor vessel containing fue.ACT(oQ b I th~+kl Restore reactor zone secondary containment integrity vithin 4 hours, or place all reactors in t least a HOT SHUTDOWN COlIDITIOH vithin the next 12 hours and in a COLD SHUT-DOMÃCO%)ITIOUS vithin the olloving 24 hours.b.ACT(o~>h>t Sc'c>uSK Pi ecHo~4~I~A t575 r,~.q,i k pseud SRs~.6.q,a,i>3.0 7>.z.AMENOMEHT No.X 74 3.7/4.7-16 BFS Unit 1 2.hftcr a secondary containment violation is determined, the standby gas trcatmcnt system vill bc operated immediately after the affected zones are isolated from the remainder of the secondary containment to confirm its ability to maintain the remainder of the secondary containment at 1/4-inch of vatcr negative pres'sure under calm vind conditions. UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE)OF LI p)3'.B.S s c 4.?.B.a Ga J.7.B.4 (Cant'd)b.Place all reactors in at least a HOT SHUTDOMH COHDITIOH vithin the next'12 hours and in a COLD SHDTDOMH COHDITIOH within the folloving 24 hours.~CC.~~~>4 C&k>6M gyp Qg g~g~~RFIV I~S'.QQ3 4.7.C.Seep da Conta e roPoM L,co 3.(i.4.2.ppplic%Lal~Q Pr3 1.Secondary containment integrity shall be maintained in the reactor zone at all times except as specified in 3.7.C.2.g g~g fg 4o AGTl6'PropS<cg Afoot 1 ko Aaml o eS 1.Secondary cantainment surveillance shall be performed as indicated belov: a.Secondary containment capability to maintain 1/4 inch of vater vacuum under calm vi.nd (c 5 mph)conditions vith a system leakage rate of not morc than 12,000 cfm, shall be demonstrated at each refueling outage prior ta refueling. If reactor zanc secondary containment integrity cannot be maintained the folloving conditions shall be met: VrspoSaD hb4'Ar~'rW Qadi~b I a.Ag<latJ b.LI p,et(~~A<8 Suspend all fuel handling operations, core altera-tions, and activities vith the potential to drain any reactor vessel containing s M~eLag tf Restore reactor zone secondary containment integrity vithin 4 hours, or place all reactors in t least a HOT SHUTDOMH COHDITIOH vithin the next 12 hours and in a COLD SHUT-DOMf COHDITIOH within the folloving 24 hours.2.After a secondary containment violation is determined, the standby gas treatment system vill be operated immediately after the affected zones are isolated from the remainder of thc secondary containment to confirm its ability to maintain the remainder of the secandary containment at 1/4-inch of vater negative pressure under calm wind conditions. See V~stll(ca4e~ 4r'CL4~pg 4r 2 Fhl I s'75 g.(, t/./+-p.s d Ses 3.<AZ.I, Z.C,g.Z.Z. BFH Unit 2 3.7/4.7-16 MEHDMENT NO.y q y pp,Q~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF NOY$8 1931 3.7.B.at st.7.B.ta d Gas eatment~S~t 3.7.B.4 (Cont'd)b.Place all reactors in at least a HOT SHUTDOWH COHDITIOH vithin the next 12 hours and in a COLD SHUTDOWH COHDITIOH vithin the folloving 24 hou See rus<4cc t)'o~fa ch nysA,~8su isrs 3.4~S.3 4.7.C.Seep da Containment t'i~gk LCP 3.fegeL 4~l, Calpi (el)1.Secondary containment integrity shall bc maintained in thc reactor zone at all times except as specified in 3.7.C.2.+LCO not applicable until jus rior to loadi fuel i o Unit 3 reac r vessel, o ded the Unit reacto pr zone not require for scconda containment nte ri units.Pro~+kg gM3*QW p>g pM I~AAo~s Secondary containment surveillance shall be performed as indicated belov: a.Secondary containment capability to maintain 1/4 inch of vater vacuum under calm vind (<5 mph)conditions vith a system inleakage rate of not more than 12,000 cfm, shall be demonstrated at each refueling outage prior to refueling. 2~If reactor zone secondary containment integrity cannot be maintained thc folloving conditions shall bc mct: Ll the potential to drain any reactor vessel containing iyl4q pg g b.cstore reactor zone secondary containment integrity vithin 4 hours, or place all reactors in at least a HOT SHUTDOWH COHDITIOH vithin thc next 12 hours and in a COLD SHUT-DOWH COHDITIOH vithin thc folloving 24 hours.QhaH~Z7 a.Suspend all fuel handling P.p>" operations, core altera-tions, and activities vith 2~After a secondary containment violation is determined, the standby gas treatment system vill be operated immediately after thc affected zones are isolated from the remainder of thc secondary containment to confirm its ability to maintain the remainder of the secondary containment at 1/4-inch of vater negative pressure under calm vind conditions. c~64's'cR$pa~~~~c$~'H 6'rs g.<,q,~MM 58~3.e.s.v.i,3.~.~.i.2. BFH Unit 3 3.7/4.7-16 NENOMENT NO.1 5 9 PAGi=>OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.2 SECONDARY CONTAINMENT ISOLATION VALVES ADNINI STRATI VE CHANGES Al A2 A3 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The current definition of Secondary Containment Integrity requires all secondary containment isolation valves (SCIVs)to be OPERABLE or in their isolation position.Thus, the current secondary containment Specification encompasses the SCIV requirements. It is proposed to provide a separate Specification for SCIVs for clarity.Thus, the new LCO will require all SCIVs to be OPERABLE, consistent with the current requirements. The applicability has been reworded to be consistent with the new definitions of MODES and to have a positive statement as to when it is applicable, not when it is not applicable. Proposed ACTIONS Note 2 (" Separate Condition entry is allowed for each penetration flow path")provides explicit instructions for proper application of the ACTIONS fo'r Technical Specification compliance. In conjunction with the proposed Specification 1.3-"Completion Times," this Note provides direction consistent with the intent of the existing Actions for inoperable isolation valves.Similarly, proposed ACTIONS Note 3 facilitates the use and understanding of the intent to consider the operability of any system affected by inoperable isolation valves and to apply applicable Actions.With the proposed LCO 3.0.6, this intent would not necessarily apply.This clarification is consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered an administrative presentation preference. BFN-UNITS I, 2,&3 Revision 0 0 A4 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.2 SECONDARY CONTAINNENT ISOLATION VALVES Amendment 159 to Unit 3 Technical Specifications added a provision to allow separating the Unit 3 reactor zone from the secondary containment envelope under certain conditions (prior to fuel loading)to expedite Unit 3 construction activities during Unit 2 operation. This provision is no longer needed and can no longer be applied.Therefore the*Note to TS 3.7.C.1 has been deleted.This change is considered administrative since it deletes a requirement that no longer applies.I 3 TECHNICAL CHANGES-NORE RESTRICTIVE Two new Surveillance Requirements have been added to ensure SCIV operability. SR 3.6.4.2.1 verifies that SCIVs isolate within the assumed times in accordance with the inservice testing program.SR 3.6.4.2.2 verifies that each SCIV actuates to its isolation position on an accident signal every 18 months.These are additional restrictions on plant operation. This change requires the movement of irradiated fuel in secondary containment and CORE ALTERATIONS to be"immediately" suspended if secondary containment is inoperable. In addition, action must be"immediately" initiated to suspend operations with the potential to drain the reactor vessel in this Condition. The current specification does not establish a time limit to suspend these activities. Immediately suspending these activities minimizes the probability of a fission product release if a reactivity event occurs while the secondary containment is inoperable. Also, immediately initiating action to suspend operation with the potential to drain the reactor vessel will minimize the potential for reactor vessel draindown and subsequent potential for fission release.Imposing a time limit to suspended these activities is a more restrictive change.TECHNICAL CHANGES-LESS RESTRICTIVE"Specific" L1 This Action has been changed to allow one valve in a penetration to be inoperable for up to 8 hours, instead of the current 4 hours.Proposed ACTION A now requires the penetration to be isolated in 8 hours.This is justified since an OPERABLE valve in the penetration remains to isolate the penetration if needed, thus the"leak tightness" of the secondary containment is still maintained. The isolated penetration is required to be verified every 31 days while a valve is inoperable, further ensuring the continued"leak tightness" of the secondary containment. Proposed ACTION B will verify that if both SCIVs in a penetration are inoperable, at least one SCIV in a penetration is closed BFN-UNITS 1, 2, 5.3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.2 SECONDARY" CONTAINMENT ISOLATION VALVES within 4 hours.This maintains consistency with the current requirements. An allowance is proposed for intermittently opening closed secondary containment isolation valves under administrative control.The allowance is presented in proposed ACTIONS Note 1, which allows opening of secondary containment penetrations on an intermittent basis for performing Surveillances, repairs, routine evolutions, etc.L2 Required Action D.l has been modified by a Note stating that LCO 3;0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.If'moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operation and the inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a,reactor shutdown.By adding an exception to LCO 3.0.3 for the failing to suspend irradiated fuel movement, an LCO 3.0.3 required reactor shutdown is avoided in MODE 1, 2, or 3.However, the plant would still be required to shutdown per proposed Required Actions C.1 and C.2 in addition to suspending fuel movement per Required Action D.l.However, this shutdown is considered less restrictive since Required Action C.1 allows the plant to be in Hot Shutdown within 12 hours versus Hot Standby within 6 hours as required by CTS 1.0.C.l.Both CTS and the proposed Required Action C.2 require the plant to be in Cold Shutdown within 36 hours.BFN-UNITS 1, 2, 5 3 Revision 0

UNIT I CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~0> o 2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at g 10K design flow on HEPh filters and charcoal adsorber banks shall show g99X, DOP removal and g99X halogenated hydrocarbon removal when tested in accordance with hHSI H510-1975. 2.a.The tests and sample analysis of Specification 3.7.B.2 shall be performed at least once per operating cycle or once every 18 months whichever occurs first for standby service or after every 720 hours of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.b.The results of laboratory carbon sample analysis shall show 290K radioactive methyl iodide removal when tested in accordance with hSTN D3803 b.Cold DOP testing shall be performed after each complete or partial replacement of the HEPT filter bank or after any structural maintenance on the system housing.c.System shall be shown to operate within ply design flow.c.Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.PAGE~uF~BPK Uait 1 3.7/4.7-14 NENIMENT KO.2 15

oscet Ac7-Prl Pn Octad o~s'mule*~ 'n i'i a h on s i ng Sg Ra KZ.I operated a ot of at least 10 hours every month.QTfohl rom aad after the date tha oae train of the standby gas treatmeat system is made or found to be faopcrablc for any reasoa, REACTOR POWER OPERATIOS and fuel handling fs permissible only durfag the succeeding 7 days unless such circuit is sooaer made OPERABLE, provfded that duriag such 7 days all active componcats of the other tvo standby gas treatment trains shall be operable.~pose)<>4 A 4e~~A o Of C.f'n~~'ir nL~lkAon Q,J v c~ao Sk'xto.9.Z. Test sealing of g ets or h using door.s pe ormc utf emic smo g er to s du ing e ch t st pe orm for co ian vi Spc i ffca ion 7.B..a k5'eci~sn~~<Once per automat c t at on of each braach of the stand<<by gas treatmeat system emo tratc from e unit'%contr ls.At least once yer manual oyerability o the byyass valve for filter cooling shall be demonstrated. c.When oae train of the s by tre tmen-ey tern be omcs yer ble the other o tr sha 1 be d oastr ed tob 0 vf 2 hours and daily ther eaf ter.BEB t 4~HClbnls C+Z Zf these coadftioas caaaot be met: a.Suspend all fuel handling oyerations, core alterations, aad activities vfth thc potential to drain~-any reactor vessel containing, fuel N 3.7/4.7-15 duffy~o c'~e~4 of lit'agiAk.d. ~aSc+bli<5>~e~sc&sdtM.Q c<rt&nned-dcrlg CoREhLT<AtTIbA, O'Ollr a Q f gpg AjbtEMDMENT NO.g 7 g Unit 1 PAGE ~~~QcTIoA s Jr.Plass a11 reactors in at least a HOT SHUTDOWN COIITIOS vithin the next 12 hours and in a COLD , SHUTDOWN CONDITION vithin the folloving 24 hours.sr'uz5AwHcn f r Q~A SFW WrS Z;e.V.t 3.7.C.4.7.C.S o 1.Secondary containment integrity shall be maintained in the reactor xone at all times except as specified in 3.7.C.2 l.Secondary containmcnt surveillance shall bc pcrformcd as indicated belov: a.Secondary containment capability to maintain 1/4 inch of vatcr vacuum under calm vind (<5 mph)conditions vith a system leakage rate of not more than 12,000 cfm, shall be demonstrated at each rcfucling outage prior to refueling. BPK Unit 1 If reactor xone secondary containment integrity cannot be maintained the folloving conditions shall bc mct: 2.hftcr a secondary containment violation is determined, the standby gas treatment system vill be operated immediately after the affected xones are isolated from the remainder of the secondary containment to confirm its ability to maintain the remainder of the secondary containment at 1/4-inch of vater negative pressure under calm vind conditions. a.Suspend all fuel handling operations, core altera-tions, and activities vith the potential to drain any r'eactor vessel containing fuel.AMENOMENT NO I 7 4 3.7/4.7-16 b.Restore reactor xone secondary containment integrity vithin 4 hours, or place all reactors in at least a HOT SHUTDOWN COHDITI01 vithin thc next 12 hours and in a COLD SHUT-DOWN COHDITION vithin thc folloving 24 hours. Cl HAR 3 0}990 l.Escape ae specified in Specification 3.7.B.3 belov, all three trains of the standby gas treatment s stem shall be'OPERABLS t all cs v ondary containment integrity is required.At least once per year, the folloving conditions shall be demonstrated. a.Pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of vater at a flov of 9000 cfm (g 10K)~&+3~+',f-<~'on/ Qi CA~4'<ZC'~lST'5 Secgo b.The inlet heaters on each circuit are tested in accordance with hBSI 8510-1975, and are capable of an output of at 1'east 40 RV c.Air distribution is uniform vithin 20X across HEPA filters and charcoal adsorbers BPK Unit 1 3.7/4.7-13 AMENDMENT NO.1.74 g.-, 5' 0' 5 C Ts Z g.F/q, 7, F FEB 1 3 1995 3.7.F.1.Thc pr ry containment p ge system hall be OPERABLE for PURGI G, except as spec fied in.7.F.2..The results of th in-place cold DOP and hal genated hydrocarbon tes s at design flovs on HEPA iltcrs and charcoal adso er banks shall show 99K DOP removal and g 99K h logenatcd hydro-carbon rem val vhen tested in accord cc vith ASSI 551-1975.b.Thc re ults of laboratory carbo.sample analysis shall shov g 85X radioactive met 1 iodide removal vh te cd in accordance vit AS D3803~c.System flov rate s 1 be shown to be vithin LOS of design flov vh tested in accordance vi ASSI 8510-1975. 2.If the provisio of 3.7.F.l.a, b~and c cannot be met, the system shall b declared inoperable. e provisions of TechnicaL Sp cification 1.C.I do not app.PURGIIC may con tfaue using the Standby Cas Treatment System.3.a.The 18-inch primary contain-ment isolation valves asso-ciatcd vith PURGIEQ may be opea during the RUN mode for a 24-hour period after entering the RUB mode and/or for a 24-hour period prior to catering the SHUTDOWH mode.The OPERABILITY of 4.7.F.~S l.At le st once every 18 mont , the prcssure drop acr ss the combined HEPA fi ters and chare al a sorber banks s 11 be emonstrated t be less than 8.5 inch of vater at system de gn flow rate (g LOX).a.Thc te s and sample analy s of Specifica-tion.7.F.1 shall be perf rmcd at least once per operating cycle or on e every 18 months, v ichever occurs first r after 720 hours of system operation and folloving significant painting, fire, or chemical release in any ventilation xo e commmicating vit the systio b.Cold DOP testi shall be performed fter each complete or artial replacement of the HEPA filter b or after any struc al mainte-nance on e system housing, c.Haiog tcd hydrocarbon test shall be per rmed after each co lete or partial re lacement of the rcoal adsorber bank o after any structural intensncc on the system housing.SC<TuSfjCac~hy~ @~g~+'Bed lS'Ts 8,v, t,g BFS Unit 1 3.7/4.7-21 AMBIOVBP gg Ej P.PAGE OF UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

3~4.2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at g 10Z design flov on HEPT filters and charcoal adsorber banks shall shov g99Z DOP removal and g99Z halogenated hydrocarbon removal vhen tested in accordance vith AHSI 8510-1975. b.The results of laboratory carbon sample analysis shall shov g90Z radioactive methyl iodide removal vhen tested in accordance vith ASTM D3803~c.System shall be shown to operate vithin glOZ design flov.2.'a.The tests and sample analysis of Specification 3.7.B.2 shall be performed at least once per operat cycle or once every 18 months whichever occurs first for standby service or after every 720 hours of system operation and folloving significant painting, fire, or chemical release in any ventilation rane communicating vith the systeme b.Cold DOP testing shall be performed after each complete or partial replacement of the HEPT filter bank or after any structural maintenance on the system housing.c.Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing+BZS Unit 2 3 7/4.7-14 PISWOMEk R+g 3 1 5 5pCcificako~ g.g.g p MAR 80 1980~A~~~Pi 4po>ci 447/aAJ Q H~R.*.'I d.Each train sha 1 be operated a total of at least 10 hours every month.....*a.-.:-/.A4 ',4'iI'yne 48'.Test sealing of gaskets for hous doors s all be pcrfo d utilizi emical sm ke genera t s during ch test per rmed.for compl cc vith Speci-ficatian 4.7.B.2.a and Specification 3.7.B.2.a. tro+SLQ Ho.Af~Rt5seo~pet~d$Q+g I Pr1pogcg gfg,latr~Aero~g, I 4.If these conditions cannot be me: Ac<I 0 tQ t" eE a.Suspend all fuel handling operations, core alterations, and activities vith the potential to drain any reactor vessel containing fue 3.Fram and after the date that one train of the standby gas treatment system is made or found to be inoperable for any reason, REACTOR POMER OPERATIOH and fuel handling is permissible only during thc succeeding 7 days un1css such circuit is sooner made OPERABLE, provided that during such 7 days all active components of the other tvo standby gas treatment trains shall bc perable.S~.C.V.R.3~a ce per W'~automatic initiatio of each branch of thc stand-by gas treatment system shall be demonstrated from unit'~nt s.sg.3.4,.gg.g /4mc nay t leas 41 anual operability of the bypass valve for filter cooling shall be demonstrated. c.Mhen one tra n o the tandby g s treatm s stem bec es inope able th other t trains sha 1 be demo trated to b OPERABLE vithin 2 hours d daily thereafter. b~.~g~ovc~f og,~;<$Q g~+ssc~klats iw~gp~,Q~*d Ca~qpg~g I-BF5 Unit 2 3.7/4.7-15 AMENDMENT ND.g 7 7 WGC g 5 ccrlicp?4a w MAR 80 1980 4 cT~aH 8 Place all reactors in at least a HOT SHUTDOMH COHDITIOH vithin the next 12 hours and in a COLD SHUTDOWNS COHDITIOH vi thin the folloving 24 hours.3.7.C.S o C t 4.7.C.Seep da Co ta 1.Secondary containment integrity shall bc maintained in the reactor zone at all times cxccpt as specified ia 3.7.C.2.1.Secondary containment surveillance shall be performed as indicated belov: a.Secondary containment capability to maintain 1/4 inch of vatcr vacuum under ca~vind (<5 mph)conditions vith a system leakage rate of not more than 12,000 cfm, shall be demonstrated at each refueling outage prior to rcfucling. 2.If reactor zone secondary containment integrity cannot be maintained thc following conditions shall be met: a.Suspend all fuel handling operations, core altcra-tioas, and activities vith thc potential to draia any reactor vessel containing fuel.b.Rcstorc reactor zone secondary containmcnt integrity vithin 4 hours, or place all reactors in at least a HOT SHUTDOMH COHDITIOH vithin thc next 12 hours aad in a COLD SHUT-DOWN COHDITIOH vithin the folloving 24 hours.2.After a secondary containment violation is determined, thc standby gas treatment system vill be operated immediately after the affected zones arc isolated from the remainder of the secondary containment to confirm its ability to maintain the remainder of the secondary containmcnt at 1/4-inch of water negative pressure under calm vind conditions. BFH Unit 2.7/4.7-16 AMENOMENT No.F77. Cl 0 MAR 80 tqso~)t'-0 36.43 r 1.Except as specified in Specification 3.7.B.3 belov, all three trains of the standby gas treatment s stem shall be OPERABLE t all times v en secondary p)containment integrity is required.4,~te 1.At least once per year, the folloving conditions shall be demonstrated. a.Pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of vater at a flov of 9000 cfm (g 10Ã).Sot', JHp44~i cog'e~4r (Q 4r Bfhf IS~gqd b.The inlet heaters on each circuit are tested in accordance vith ANSI N510-1975, and are capable of an output of at least 40 kM.c.Air distribution is uniform vithin 20K across HEPA filters-and charcoal adsorbers. Troppo'~Sg S.$,6,3.2.BFH Unit 2 3.7/4.7-13 AMENDMENT ND.X'7 T eGc 6 oF~

.7.F.BFK-Unit 2 1.The primary contaiame purge system shall be OPE LE for PURGIHG, except as pecified in 3.7.F.2.a.The results f the in-place cold DOP d halogenated hydrocar n tests at design flovs o HEPA filters and chare 1 adsorber banks sha shov g 99K DOP removal an g 99K, halogenated hydro-rbon removal when tested n accordance vith AHSI H510-1975~b.The results of laborato carbon sample analysi shall shov g 85K radioact e methyl iodide remo 1 vhen tested in accord ce vith ASTM D3803.c.System flo rate shall be showa to e within g 10K of desi flov vhen tcstcd in accordance vith AHSI H510-1975~If the provisions of 3.7.F.l.a, b, and c cannot be met, the system shall be declared inoperable. The provisions of Technical Specification 1.C.1 do not apply.PURGIHG may coa-tiaue using the Standby Gas Treatment System.~a.e-ch pr ry conta ment isolation valves asso-ciated vith PURGIHG may be open during, the RUH mode for a 24-hour period after entering the RUH mode and/or for a 24<<hour period prior to entering the SHUTDOMH mode.The OPERABILITY of..3.7/4.7-21 4.7.F.SZRttell l.At least once every 18 mon , the pressure drop acr ss the combined HEPA f ters and charcoal orber banks shall be demonstrated to be lcss-,'.-than 8.5 inches of vatei at system design flov rate (g 10K).a.The tests and sample analysis of Specifica-tion 3.7.F.1 11 be performed at east onc pcr operati cycle or once eve 18 months, vhichevc occurs first or aft 720 hours of syst operation and fo ovtng significant p ating, fire, or emical release in any ventilation zone comanmicating vith the system.b.Cold DOP testing shall be performed aftc each complete or par al replacement of e HEPA~filter bank after any structu al mainte-nance on e system" housing.c.Halog tcd hydrocarboa tes shall be pcr ormed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.~<~HsfICicak'so~ Qr t:h~gqg 4~am i~~~~I~gg@O~Ny.pg I~1 Vr.

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP u PCCggiC'a Z, lo+>3 0 1990 l.Except as specified in Specification 3.7.B.3 belov,+g~<<>all three trains of the'standby gas treatment s stem shall be OPERABLE at all times v en secondary containment integrity is required.At least once per year, the folloving conditions shall be demonstrated. a.Pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of vater at a flov of 9000 cfm (g 10K).ac gusfigcahon foe Cho~gcs P r 5~t 5 Ts 5ccvi~b.The inlet heaters on each circuit are tested in accordance vith AHSI H510-1975, and are capable of an output of at least 40 kM.c.Air distribution is uniform vithin 20K across HEPA filters and charcoal adsorbers. Z y~eP Se s.e.q.z,~BFI unit 3 3.7/4.7-13 AMENDMENT Na.X4$PAGE A 0'

2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at g 10Z design flow on HEPA filters and charcoal adsorber banks shall show g99Z DOP removal and g99Z halogenated hydrocarbon removal when tested in accordance with AHSI H510-1975. 2.a.e tests and sample analysis of Specification 3.7.B.2 shall be performed at least once per operating cycle or once every 18 months whichever occurs first for standby service or after every 720 hours of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.b.The results of laboratory carbon sample analysis shall show g90Z radioactive methyl iodide removal when tested in accordance with ASTM D3803.b.Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing.c.System shall be shown to operate within glOX design flow.c.Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.BPH Unit 3 3.7/4.7-14 NENO<<Ee NO I 8 g CO MME S S MS 5f'ec>f'cuban 3.C.Q 3~3 0 I90 fAePOSeg/CA'on D~~SR3<.9.3',l/N2.operated ota of at least 10 hours every month.On~at&4 or 5i~MlR in'god~ianna)3.From and after the date that one train of the standby gas treatment system is made or found to be inoperable for any reason, REACTOR POWER OPERATIOH and fuel handling is permissible only during thc succeeding 7 days unless such circuit is sooner made OPERABLE, provided that during such 7 days all active components of thc other tvo standby gas treatment trains shall be operable.iloys~A Nog g gr~,,M 46 n~04 C4E~I~z WHonC I 4.If these conditions cannot bc mct: C~onc train of the st dby g s tre ment sys em bec mes i perable the ther t o tra s shall be d nstra ed to be PERAB vithx 2 hours dail hereafter. e.Test sealing of gaskets r ho ing do s shall be pcrfo ed uti izing ch ical moke ge era-tors duri each t st perfo ed f r compli ce v th Spec ficatio 4.7..2.a and Specification .7.B.2.a.5g 3'.6 v h a.ce pcr automatic initiation of each branch of t e stand-by gas treatment system al e emoqs tra~rom ca unit&contro 5R~.<.V.~V~eoe&f~*manual operability o the bypass valve for filter cooling shall be demonstrated. BFH Unit 3<Ihg IH~a oaf 4 lrrRdi'afcd ~en bligh ih Wc scco~y co&;hnehg~<'"$~<F ift,7FRtl TXCWS, Or ggri~~OT't Rlt's a.Suspend all fuel Rtbh'andling operations, C.>E core alterations, and activities vith the potential to drain any reactor vessel containing fue s~Miale Al lq 7-15 AMENDMENT No.74 g 3.7/4. 4 RChon 6 b.Place all reactors in at least a HOT SHUTDOWH COHDITIOH within the next 12 hours and in a COLD SHUTDOWH COHDITIOH within the following 24 hours.5'uSfjke~+, (~~peg 4<~n)/5 75 3.4.Q.I 3.7.C.Seconda o a nme 4.7.C.Seep da Containment

• 1.Secondary containment integrity shall be maintained in the reactor zone at all times except as specified in 3.7.C.2.1.Secondary containment surveillance shall be performed as indicated below:.*LCO not applicable until)ust prior to loading fuel into the Unit 3 reactor vessel, provided the Unit 3 reactor zone is not required for secondary containment integrity for other units.a.Secondary containment capability to maintain 1/4 inch of water vacuum under calm wind (<5 mph)conditions with a system inleakage rate of not more than 12,000 cfm, shall be demonstrated at each refueling outage prior to refueling.

BFH Unit 3 2.If reactor zone secondary containment integrity cannot be maintained the following conditions shall be met: a.Suspend all fuel handling operations, core altera-tions, and activities with the potential to drain any reactor vessel containing fuel.b.Restore reactor zone secondary containment integrity within 4 hours, or place all reactors in at least a HOT SHUTDOWH COHDITIOH within the next 12 hours and in a COLD SHUT-DOWH COHDITIOH within the following 24 hours.2.After a secondary containment violation is determined, the standby gas treatment system will be operated immediately after the affected zones are isolated from the-remainder of the secondary containment to confirm its ability to maintain the remainder of the secondary containment at 1/4-inch of water negative pressure under calm wind conditions. ENDMEHT NO.159 was~op~

3.7.F.4.7.F.BPK Unit S 1.The primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3.7.F.2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall shov g 99K DOP removal and g 99%halogenated hydro-carbon removal vhen tcstcd in accordance vith AHSI H510-1975. b.The results of laboratory carbon sample analysis shall shov g 85K radioactive methyl iodide removal vhcn tested in accordance vith AS'3803.c.System flov rate shall be sham to be vithin g 10X of design flov vhen tested in accordance vith AHSI 5510-1975'~If the provisions of 3.7.P,l.a~ b, and c cannot be met, the system shall be declared inoperable. The provisions of Technical Specification 1.C.1 do not apply.PURCINC may con>>tinue using the Standby Caa Treatment System.3.a.The 18-inch primary contain-ment isolatian valves asso-ciated vith PURGIHG may be open during the RUH mode f'r a'4-hour period after entering the RUE mode and/or for a 24-hour period prior to catering the SHUTDOWI mode.The OPERABILITY of 3.7/4.7-21 1.At least once every 18 months, the pressure dro across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to bc less than 8.5 inches of vater at system design flov rate (g lOZ).a.The tests and sample analysis of Specifica-tion 3.7.F.l shall be performed at least once per operating cycle or once every 18 months, whichever occurs first or after 720 hours of system operation and folloving significant painting, fire, or chemical release in any ventilation zone coamunicating vith the systems b.Cold DOP testing shall be performed after ea complete or partial replacement of thc HEP filter bank or after any structural mainte-nance an the system housing.c.Halogenated bydrocarbo testing shall be perfarmed after each complete ar partial replacement of the charcoal adsorber bank or after any atructura maintenance on the system housing.%is QPgqgp~4 c~+~$w~~~<~S4 I'5 amo~ur s0.EBS\I

t JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.3-STANDBY GAS TREATMENT SYSTEM ADMINISTRATIVE CHANGES Al A2 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The technical content of this requirement is being moved to Section 5.0 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433.Any technical changes to this requirement will be addressed within the content of proposed Specification 5.5.7.A surveillance requirement (proposed.SR 3.6.4.3.2) is added to clarify that the tests of the Ventilation Filter Testing Program must also be completed and passed for determining operability of the SGT System.Since this is a presentation preference that maintains current requirements, this change is considered administrative. A3 The description of the signal used to automatically initiate the SGT System"actual or simulated initiation signal" has been added for clarity.This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and no change is intended.A new ACTION is proposed (ACTION D)which directs entry into LCO 3.0.3 if two or more required standby gas treatment subsystems are inoperable in Modes 1, 2, or 3.This avoids confusion as to the proper action if in Modes 1, 2, or 3 and simultaneously handling fuel, conducting CORE ALTERATIONS, or operations with the potential for draining the reactor vessel.Since the proposed ACTION effectively results in the same action as the current specification, this change is considered administrative. A5 The Frequency for verifying SGTS automatic initiation has been changed to 18 months from once per operating cycle.The BFN operating cycle is currently defined as 18 months.As such this is a change in presentation only and is therefore administrative. BFN-UNITS 1, 2, 5 3 Revision 0 PAcs I or~ JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.3-STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGES-MORE RESTRICTIVE This change requires the movement of irradiated fuel in secondary containment and CORE ALTERATIONS to be"Immediately" suspended if secondary containment is inoperable. In addition, action must be"Immediately" initiated to suspend operations with the potential to drain the reactor vessel in this Condition. The current specification does not establish a time limit to suspend these activities. Immediately suspending these activities minimizes the probability of a fission product release if a reactivity event occurs while the secondary containment is inoperable. Also, immediately initiating action to suspend operation with the potential to drain the reactor vessel will minimize the potential for.reactor vessel draindown and subsequent potential for fission release.Imposing a time limit to suspended these activities is a more restrictive change.CTS 4.7.B.2.d requires each train to be operated a total of at least 10 hours each month.Proposed SR 3.6.4.3.1 requires each train to be.operated continuously for 10 hours.As such, the proposed SR is considered more restrictive. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1 LA2 Details on methods of testing gasket seals for housing doors has been deleted.This type of detail will be retained in plant procedures and/or system operating instructions. Details on the method of performing Standby Gas Treatment system surveillance requirements have been relocated to plant procedures. Changes to the procedure will be controlled by licensee controlled programs."Specific" L1 The proposed change will delete the requirement to test the other SGT subsystems when one subsystem is inoperable. The requirement for demonstrating operability of the redundant subsystems was originally chosen because there was a lack of plant operating history and a lack of sufficient equipment failure data.Since that time, plant operating experience has demonstrated that testing of the redundant subsystems when one subsystem is inoperable is not necessary to provide adequate assurance of system operability. This change will allow credit to be taken for normal periodic surveillances as a demonstration of operability and availability of the BFN-UNITS 1, 2, 5 3 Revision 0 PAGE JUSTIFICATION FOR CHANGES BFN ISTS 3.6.4.3-STANDBY GAS TREATMENT SYSTEM L2 L3 remaining components. The periodic frequencies specified to demonstrate operability of the remaining components have been shown to be adequate to ensure equipment operability. As stated in NRC Generic Letter 87-09,"It is overly conservative to assume that systems or components are inoperable when a surveillance requirement has not been performed. The opposite is in fact the case;the vast majority of surveillances demonstrate the systems or components in fact are operable." Therefore, reliance on the specified surveillance intervals does not result in a reduced level of confidence concerning the equipment availability. Also, the current Standard Technical Specifications (STS)and, more specifically, all the Technical Specifications approved for recently licensed BWRs accept the philosophy of system operability based on satisfactory performance of monthly, quarterly, refueling interval, post maintenance or other specified performance tests without requiring additional testing when another system is inoperable (except for diesel generator testing, which is not being changed).An alternative is proposed to suspending operations if a SGT subsystem cannot be returned to OPERABLE status within seven days, and movement of irradiated fuel assemblies, CORE ALTERATIONS, or operations with the potential for draining the reactor vessel are being conducted. The alternative is to initiate two OPERABLE subsystems of SGT and continue to conduct the operations. Since two subsystems are sufficient for any accident, the risk of failure of the subsystems to perform their intended function is significantly reduced if they are running.This alternative is less restrictive than the existing requirement. However, the proposed alternative ensures that the remaining subsystems are Operable, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.This change is consistent with NUREG-1433. The Required Actions of C and E.I have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.If moving irradiated fuel assemblies while in MODE I, 2, or 3, the fuel movement is independent of reactor operation and the inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.By adding an exception to LCO 3.0.3 for the failing to suspend irradiated fuel movement, an LCO 3.0.3 required reactor shutdown is avoided in MODE I, 2, or 3.However, the plant would still be required to shutdown per proposed Required Actions B.1 and B.2 in addition to suspending fuel movement per Required Actions C.I and E.l.However, this shutdown is considered less restrictive since Required Action B.1 allows the plant to be in Hot Shutdown within 12 hours versus Hot Standby within 6 hours as required by CTS 1.0.C.l.Both CTS and the proposed Required Action B.2 require the plant to be in Cold Shutdown within 36 hours.BFN-UNITS I, 2, 8.3 Revision 0 PAGE~OF~ t0 JUSTIFICATION FOR CHANGES CTS 3.7.F/4.7.F -PRINRY CONTAINMENT PURGE SYSTEN RELOCATED CHANGES Rl CTS.3.7.F.1&2 and 4.7.F requirements have been relocated to the Technical Requirements Manual (TRM).The Primary Containment Purge System is normally isolated and normally not required to be functional during power operation. It does provide the preferred exhaust path for purging the primary containment; however, the SGTS can be used to perform the equivalent function.The supply and isolation valves are depended on to function properly for containment isolation, which is covered in proposed BFN ISTS Section 3.6.1.3, Primary Containment Isolation Valves.BFN-UNITS 1, 2,&3 Revision 0

Enclosure III Volume 7 'l TABLE OF CONTENTS Section 8 8 8 8 ,8 Ba 3.1 3.1.1 3.1~2 3.1.3 3.1.4 3.1.5 3.'.6 3.1.7 3.1.8 8 2.0 8 2.1.1 8 2.1.2 8 3.0 8 3.0 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGZN (SDM).......~Reactivity Anomalies Control Rod OPERABILITY Control Rod Scram Times Control Rod Scram Accumulators Rod Pattern Control Standby Liguid Control (SLC)System Scram Discharge Volume (SDV)Vent and Dra Valves~~in SAFETY LIMITS (SLs)Reactor Core SLs Reactor Coolant System (RCS)Pressure SL LIMZTZNG CONDITION FOR OPERATION (LCO)APPLICABILITY SURVEILLANCE REQUIREMENT (SR)APPLICABILITY ~Pa r~o 8 2.0 X.,-8 2.0 8 2.0.~7 830"1~'8"3'.0-1P 8 3,1 3, 8 3.1-1 8 3.>>0.'8'.1 13.'"'8 3~I-?2 8 3>>1" 29..8'".3.-34' 3'"-39'3.1-46 8 8 3.2 3.2.1 8.3~2.2 8 3.2.3 8 3.2.4 8 8 8 8 8 3~3~1'.2 3.3.2.1 3.3.2.2 3.3.3.1.8 8 3.3.3.2 3.3.4.1 B 3.3.4.2 3.3.5.1 3.3.5.2 3.3.6.1 3.3.6.2 3.3.7.1 8 8 3.3.8~1 3.3.8.2 3.3 3.3.1.1 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)MINIMUM CRITICAL POWER RATIO (MCPR)LZNEAR HEAT GENERATION RATE (LHGR)Average Power Range Monitor (APRM)Gain and Setpoints INSTRUMENTATION Reactor Protection System (RPS)Instrumentation Source Range Monitor (SRM)Znstrumentation Control Rod Block Instrumentation Feedwater and Main Turbine Trip Instrumentation Post Accident Monitoring (PAM)Instrumentation Backup Control System End of Cycle Recirculation Pump Trip (EOC-RPT)Instrumentation Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation Emergency Core Cooling System (ECCS)Instrumentation Reactor Core Isolation Cooling (RCIC)System Instrumentation Primary Containment Isolation Instrumentation Secondary Containment Isolation Instrumentation .Control Room Emergency Ventilation (CREV)System Instrumentation Loss of Power (LOP)Instrumentation Reactor Protection System (RPS)Electric Power Monitoring 8 3.2-1 8 3.2-1'3'-4 8 3.2-9 8;3~2 12'8 3.3-1 8 3.3-1 8"3.3 33'8 3.3.~4?8 3.3-53 8 3.3-60 8 3'-,71 8 3.3-80 8 3.3-89 8 3.3-98 8 3.3-135 8 3.3-143 8 3.3-165 8 3.3-176 8 3.3-188 8 3.3-196 BFN Unit 2 .1 ili Section~Pa e No.B B B B B B B B B B B B B B B B'B B B B'B B B B B B B B B B B B B B 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3'.8 3.4.9 3.5 3.5'3.5.2 3.5.3 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3~6.1.6 3.6.2.1 3.6.2.2 3.6.2.3 3.6.2.4 3.6.2.5 3.6.2.6 3.6.3.1 3.6.3.2 3.6.4'3.6.4.2 3.6.4.3 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3'.7.6 3.7.7 3.7'REACTOR COOLANT SYSTEM (RCS)..~Recirculation Loops Operating Jet Pumps Safety/Relief Valves (S/RVs)RCS Operational LEAKAGE RCS Leakage Detection Instrumentation RCS Specific Activity Residual Heat Removal (RHR)Shutdown Cooling System-Hot Shutdown Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown RCS Pressure and Temperature (P/T)Limits EMERGENCY CORE COOLING'SYSTEMS (ECCS)AND REACTOR CORE ISOLATION COOLING (RCIC)SYSTEM ECCS-Operating ECCS-Shutdown~.~~.~~~RCIC System CONTAINMENT SYSTEMS Primary Containment Primary Containment Air Lock Primary Containment Isolation Valves (PCIVs)Drywell Air Temperature Reactor Building-to-Suppression Chamber Vacuum Breakers Suppression Chamber-to-Drywell Vacuum Breakers Suppression Pool Average Temperature Suppression Pool Water Level Residual Heat Removal (RHR)Suppression Pool Cooling o~t~~~~~~~~~~~~~~~~Residual Heat Removal (RHR)Suppression Pool'Spray Residual Heat Removal (RHR)Drywel1 Spray Drywell-to-Suppression Chamber Differential Pressure e~~~~~~~~~~I'~~~~~Containment Atmosphere Dilution (CAD)System Primary Containment Oxygen Concentration Secondary Containment Secondary Containment Isolation Valves ('SCIVs)Standby Gas Treatment (SGT)System PLANT SYSTEMS.........~~~~~~~~~~Residual Heat Removal Service Water (RHRSW)System i e~~~~~~~~~~~~Emergency Equipment Cooling Water (EECW)System and Ultimate Heat Sink (UHS)Control Room Emergency Ventilation (CREV)S ystem Control Room Air Conditioning (AC)System Main Condenser Offgas Main Turbine Bypass System Spent Fuel Storage Pool Water Level B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B B 3.4-1 3.4-1 3.4-9 3.4-14 3.4-19 3.4-25 3.4-31 3'-35 3'-40 3.4-45 3.5-1 3.5-1 3.5-18 3.5-24 3'-1 3'-1 3.6-6 3.6-14 3.6-28 3.6-31 3.6-37 3.6-43 3.6-49 3.6-52 3.6-57 3.6-75 3.6-67 3.6-70 3.6-75 3.6-78 3.6-83 3.6-89 3~7 1 3.7-1 3~7 7 3.7-12 3.7-19 3.7-30 3.7-24 3.7-28 BFN Unit 2 4 ill~I>ili ~Sectio Pacae No.B 3.8 B'.8.1 B 3.8.'2 B 3.8.3 B 3.8.4 B 3.8.5 B 3.8.6 B 3.8.7 B 3.8.8, ELECTRICAL POWER SYSTEMS AC Sources-Operating AC Sources-'Shutdown Diesel Fuel Oil, Lube Oil, DC Sources-Operating DC Sources-'Shutdown Battery Cell'Parameters Inverters-Operating Inverters-Shutdown~~~~~~~~~and Starting Air B B B B B B B B~B 3.8-1 3'.8-1 3.8-28 3.8-35 3.8-42 3'.8-51 3.8-55 3.8-62 3.8-73 B.3.9 B 3.9.1 B 3.9.2 B 3.9.'3.B.3.9.4 ,B'.9.5 B 3~9~6 ,B 3.9.7'3~9.8.REFUELING OPERATIONS Refueling. Equipment Interlocks Refuel Position One-Rod-Out Interlo'ck Control Rod Position Control Rod Position.Indication Control Rod OPERABILITY -Refueling Reactor Pressure Vessel (RPV)Water Level Residual Heat Removal;(RHR) -High Water Level Residual Heat Removal (RHR)-Low Water Level,.B B: B B: B B B B B 3.'9-1 3.9-1 3.9-5 3.9-9 3.9-12 3.9-16 3'-19 3.9-22.3.9-26.B 3.10 B 3~10~1 B 3.10.2 B 3.10.3 B 3.10.4 B 3.10.5'B 3.10.6 B 3.10.'7 B 3.10.8 SPECIAL'OPERATIONS ~~~~~~~~~~~~~.~~Inservice Leak and Hydrostatic Testing Operation Reactor Mode, Switch Interlock Testing Single Control Rod Withdrawal' Hot Shutdown Single Control Rod'Withdrawal -Cold Shutdown Single Control Rod Drive (CRD)Removal-Refueling Multiple Control Rod,Withdrawal-Refueling Control Rod'esting-Operating SHUTDOWN MARGIN (SDM)Test-Refueling B 3.10-1 B B B B 3.10-21 3.10-26 3'.10-29 3'.10-33'B 3.10-1 B 3.10-6 B 3.10-11 B 3.10-16 BFN.Unit 2 g hduncanhulbasa s.toe /J~5)ill 4i Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)8 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1)requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients. The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated.Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the HCPR is not less than the limit specified in Specification 2.1.1.2 for General Electric Company (GE)fuel.MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs.The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions. Mhile fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused c'/adding perforations signal a threshold beyond which still greater thermal stresses may cause gross,'rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (,i.e., NCPR=1.00).These conditions represent a significant departure from the condition intended by design for planned operation. The NCPR fuel cladding integrity SL ensures that during normal operation and during abnormal operational transients, at least 99.9%of the fuel rods in the core do not experience transition boiling.{continued) BFN-UNIT 2 B 2.0-1 Amendment il~i~II Reactor Core SLs B 2.

1.1 BACKGROUND

(continued) Operation above the boundary of the nucleate boiling regime could result in excessive c1adding temperature because of the onset of transition boiling 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.APPL ICABL'E SAFETY ANALYSES The fuel cladding, must not sustain damage as a result of normal operation and abnormal operational transients. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9%of the fuel rods in the core would not'be expected to experience the onset of transition boiling.The Reactor Protection System setpoints (LCO 3.3.1.1,"Reactor Protection System (RPS)Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor'oolant System water level, pressure, and THERMAL POWER level that.woul'd result in reaching the MCPR limit.2.1.1.1 Fuel Claddin Inte rit GE critical power correlations are applicable for all critical power calculations at pressures~785 psig and core flows~10%of rated flow.For operation at low pressures or low flows, another basis is used, as follows: The static head across the fuel bundles due onl'y to elevation effects from liquid only in the channel, core bypass region, and annulus at zero power, zero flow is approximately 4.5 psi.At all operating conditions, this pressure differential is maintained by the bypass region of the core and the annulus region of the vessel.The elevation head provided by the annulus produces natural circulation flow conditions which have balancing pressure head and loss terms (continued) BFN-UNIT 2 B 2.0-2 Amendment 0 0 Reactor Core SLs B 2.1.1 BASES inside the core shroud.This natural circulation principle maintains a core plenum to plenum pressure drop of about 4.5 to 5 psid along the natural circulation flow line of the P/F operating map.In the range of, power levels of interest, approaching 25%of rated power below which thermal margin monitoring is not required, the pressure drop and density head terms tradeoff for power changes such that natural circulation flow is nearly independent of reactor power.This characteristic is represented by the nearly vertical portion of the natural circulation line on the P/F operating map.Analysis has shown that the hot channel flow rate is)28,000 lb/hr in the region of operation with power-25%and core pressure drop of about 4.5 to 5 ps,id.Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at 28,000 lb/hr is approximately 3 NWt.With the design peaking factors, this corresponds to a core thermal power of more than 50%.Thus operation up to'25%of rated power with normal natural circulation available. is conservatively acceptable even if reactor pressure is equal to or below 800 psia (the limit of the range of applicability of GETAB/GEXL for GE fuel).If reactor power is-significantly less than 25%of rated (e.g., below 10%of rated), the core flow and the channel flow supported by the available driving head may be less than 28,000 lb/hr (along the lower portion of the natural circulation flow characteristic on the P/F map).However, the critical power that can be supported by the core and hot channel flow with normal natural circulation paths available remains well above the actual power conditions. The.inherent characteristics of BWR natural circulation make power and core flow follow the natural circulation line as long as normal water level is maintained.(continued) BFN-UNIT 2 B 2.0-3 Amendment ik~il~ Reactor Core SLs 8 2.1.1 APPLICABLE SAFETY ANALYSES 2.l.l.I Fuel Claddin Inte rit (continued) Thus, operation with core thermal power below 25%of rated without thermal, margin surveillance is conservatively acceptable even for reactor operations at natural circulation. Adequate fuel thermal margins are also maintained without further surveillance for the low power conditions that would be present if core natural circulation is below 10%of rated flow (the limit of applicability of the GETAB/GEXL correlations for GE fuel).2.1.1.2 HCPR The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated.Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur.Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.However, the uncertainties in monitoring the core operating state and in the procedures used to cal'culate the critical power result in an uncertainty in the value of the critical power.Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9%of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties. The HCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power.The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2.Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the NCPR SL statistical analysis.(continued) BFN-UNIT 2 B 2.0-4 Amendment il'I 0 Reactor Core SLs 8 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) 2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat.If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced.This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes (2/3 of the core height.The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.SL 2.1.l.1 and SL 2.1..1.2 ensure that the core operates within the fuel design criteria.SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations. APPLICABILITY SLs 2.l.1.1, 2.1.1.2 and 2.1.1.3 are applicable in all MODES.SAFETY LIMIT Y IOLAT ION S Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100,"Reactor Site Criteria," limits (Ref.3).Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours.The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.(continued) BFN-UNIT 2 B 2.0-5 Amendment il~O~ Reactor Core SLs B 2.1.1'BASES (continued) REFERENCES l.10 CFR 50, Appendix A, GDC 10.2.GE SIL No.516, Supplement 2, January 19, 1996.3.10 CFR 100.BFN-UNIT 2 B 2.0-6:Amendment 0 RCS Pressure SL 8 2.1.2 B 2.0 SAFETY LIHITS (SLs)B 2.1.2'Reactor Coolant System (RCS)Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects 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. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. Per 10 CFR.50, Appendix A, GDC 14,"Reactor Coolant Pressure Boundary," and GDC 15,"Reactor Coolant System Design" (Ref.1), the reactor coolant pressure boundary (RCPB)shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and abnormal operational transients. During normal operation and abnormal operational transients, RCS pressure is limited from exceeding the design pressure by more than'10%, in accordance with Section III of the ASHE.Code (Ref.2).To ensure system integrity, all RCS components are hydrostatically tested at 125%of design pressure, in accordance with ASHE Code requirements, prior to initial operation when there is no fuel in the core.Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1-,"Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASHE Code, Section XI (Ref.3).Overpressurization of the RCS could result in a breach of the RCPB reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100,"Reactor Site Criteria" (Ref.4).If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere. BFN-UNIT 2 B 2.0-7 (continued) Amendment il il RCS Pressure SL B 2.1.2 BASES (continued) APPLICABLE SAFETY ANALYSES The RCS safety/relief valves and the Reactor Protection System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded.The RCS pressure SL has been selected such that it is at a ,pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASHE, Boiler and Pressure Vessel Code, 1965 Edi,tion, including Addenda through the summer of 1965 (Ref.5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.The SL of 1325 psig, as measured in the reactor steam dome is equival'ent to 1375 psig at the lowest elevation of the RCS.The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1, 1967 Edition (Ref.6), and the additional requirements of GE design and procurement specifications (Ref.7)which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10, for the reactor recirculation piping, which permits a maximum pressure transient of 120%of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping.The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASHE Code, Section III, is 110%of design pressure.The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120%.of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping.Mhen the 20%.allowance (230 and 265 psig)allowed by USAS Piping Code, Section B31.1, for pressure transients is added to the design pressures, transient pressure limits of 1378 and 1591 psig are established. The most limiting of these allowances is the 110%of design pressure for the RCS pressure vessel;therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.APPLICABILITY SL 2.1.2 applies in all NODES.BFN-UNIT 2 B 2.0-8 (continued) Amendment ik~O~ RCS Pressure SL B 2.1.2 BASES (continued) SAFETY LIHIT VIOL'AT I ONS 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).Therefore,, it is required to insert.all insertable control rods and restore compliance with the SL within 2 hours.The 2 hour Completion Time ensures that the operators take prompt.remedial action and also assures that the probabil.i.ty of an accident occurring during this period is minimal.REFERENCES 0 l.10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.2.ASHE, Boiler and Pressure Vessel Code, Section III, Article NB-7000.3.ASHE, Boiler and Pressure Vessel Code, Section XI, Article IW-5000.'4.10 CFR 100.5.ASHE, Boiler and Pressure Vessel Code,:Section III, 1965 Edition, Summer of 1965 Addenda.6.ASHE, USAS, Nuclear Power Piping Code, Section B31.1, 1967 Edition., 7.BFN General Electric Design Specification 22A1406,"Pressure.Integrity of Piping and Equipment Pressure.Parts," Revision 2, April 28, 1970.BFN-UNIT 2 B 2.0-9 Amendment ~i il~ LCO Applicability B 3.0 BASES LCOs LCO 3.0.1 through LCO 3.0.7 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 Completion Times constitutes compliance with a Specification; and b.Completion of the Required Actions is not required when an LCO is met 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 (continued) BFN-UNIT 2 B 3.0-1 Amendment ~i i LCO Applicability B 3.0 BASES LCO 3.0.2 (continued) 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. 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 3.4.9,"RCS Pressure and Temperature 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. Alternatives that would not result in redundant equipment being inoperable should be used instead.Doing so limits the time both subsystems/divisions of a, safety function are inoperable and limits the time other conditions exist which 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 (continued) BFN-UNIT.2 B 3.0-2 Amendment

LCO Applicability B 3.0 LCO 3.0.2 (continued) 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 and no other Condition applies;or b.The condition of the uni't is not specifically addressed by the associated 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. This Specification delineates the time limits for placing the unit in a safe NODE 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, I 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 NODES 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 (continued) BFN-UNIT 2 B 3.0-3 Amendment iS~il' LCO Applicability B 3.0 BASES LCO 3.0.3 (continued) 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 been performed. c.ACTIONS exist that do not have expired Completion Times.These Completion 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 Specification

3.0.3 allow

37 hours for the unit to be in NODE 4 when a shutdown is required during NODE 1 operation. If the unit is in a lower NODE of operation when a shutdown is required, the time limit for reaching the next lower NODE applies.If a lower NODE is reached in less time than allowed, however, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced.For example, if NODE 2 is reached in 2 hours, then the time allowed for reaching NODE 3 is the next 11 hours, because the total time for reaching NODE 3 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 NODES 1, 2, and 3, 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 4 and 5 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 NODE 1, 2, or 3)because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.(continued) BFN-UNIT 2 B 3.0-4 Amendment ~~il' LCO Applicability B 3.0 LCO 3.0.3 (continued) 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 3.7.6,"Spent Fuel Storage Pool Mater Level." LCO 3.7.6 has an Applicability of"During movement of irradiated fuel assemblies in.the spent fuel storage pool." Therefore, this LCO can be applicable in any or all MODES.If the LCO and the Required Actions of LCO 3.7.6 are, not met while in MODE I, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.6 of"Suspend movement of irradiated fuel assemblies in the spent 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.'t precludes placing the unit in a different MODE or other specified condition stated in that Applicability (e.g., Applicability desired to be entered)when the following exist: a.Unit conditions are such that requirements of the LCO would not be met in the Applicability desired to be entered;and b.Continued noncompliance with the LCO requirements, if the Applicability wer e entered, would result in the unit being required to exit the Applicability desired to be entered to comply with the Required Actions.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.The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good (continued) BFN-UNIT 2'B 3.0-5 Amendment ik~' LCO Applicability B 3.0 LCO 3.0.4 (continued) practice of restoring systems or components to OPERABLE status before unit startup.The provisions of L'CO 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.Exceptions to LCO 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification. LCO 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4;or MODE 1 from MODE 2.Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3.The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3)because the ACTIONS of individual specifications sufficiently define the remedial measures to be taken.[In some cases (e.g ,.)these ACTIONS provide a Note that states"awhile this LCO is not met, entry into a MODE or other specified, condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] 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, changing'MODES or other specified conditions whil'e in an ACTIONS Condition, either in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated 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.BFN-UNIT 2 B 3.0-6 (continued) Amendment 0 LCO Applicability B 3.0 BASES.(continued) 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 testing to demonstrate: a.The OPERABILITY of the equipment being returned to service;or b.The OPERABILITY of other equipment; or c.That variables are within limits.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 allowed testing.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 SRs.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 an SR 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'ndicate the appropriate response during the performance of an SR on another channel in the same trip system.LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 when a supported system LCO is not met solely due to a support system LCO not being met.This exception is provided because LCO 3.0.2 would require that the Conditions and (continued) BFN-UNIT 2.B 3.0-7 Amendment il~il LCO Applicability B 3.0 LCO 3.0.6 (continued) 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 plant is maintained in a safe condition are specified in the support systems'CO's Required Actions.The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems'CO's Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the plant 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 immedi'ately 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 5.5.11,"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 3.0.6.1 The SFDP requires cross division checks to identify a loss of safety function for those support systems that support safety systems are required.The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.If this 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.BFN-UNIT 2 B 3.0-8 (continued) Amendment ~~ll'I LCO Applicability B 3.0 BASES (continued) 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. Special Operations LCOs in Section 3.10 allow specified 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 NODE 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 Special Operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS.Compliance with Special Operations LCOs is optional.A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be followed.Mhen a Special Operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCO's Applicability (i.e., should the requirements of this other LCO not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO).However, there are instances where the Special Operations LCO's ACTIONS may direct the other LCO's ACTIONS be met.The Surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO.If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs)are required to be met concurrent with the requirements of the Special Operations LCO.BFN-UNIT 2 B 3.0-9 Amendment il'l SR Applicability B 3.0 B 3.0 SURVEILLANCE RE(UIREMENT (SR)APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to al.l 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 NODES 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.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, although still meeting the SRs;or j b.The requirements of the Surveillance(s) are known to be not met between.required Surveillance performances. Surveillances do not have to be performed when the unit is in a NODE or other specifi'ed condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification. 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.(continued) BFN-UNIT 2 B 3.0-10'mendment ~, il SR Applicability B 3.0 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 NODE 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 NODE or other specified condition where other necessary post maintenance tests can be completed. Some examples of this process are: a~Control Rod Drive maintenance during refueling that requires scram testing at)800 psi.However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.4.3 is.satisfied, the control rod can be considered OPERABLE.This allows startup to proceed to reach 800 psi to perform other necessary testing.b.High pressure coolant injection (HPCI)maintenance during shutdown that requires system functional tests at a specified pressure.Provided other appropriate testing is satisfactorily completed, startup can proceed with HPCI 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 25K 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., (continued) BFN-UNIT 2 B 3.0-11 Amendment 0 Il SR Applicability B 3.0 BASES SR 3.0.2 (continued) 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 25/.extension of the interval specified in the Frequency does not apply.These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of"in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the TS.The TS cannot in and of themselves extend a test interval.specified in the regulations. Therefore, there is a Note in the Frequency stating,"SR 3.0.2 is not applicable." 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 (continued) BFN-UNIT 2 B 3.0-12 Amendment il~il~ SR Applicability B 3.0 SR 3.0.3 (continued) 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 less, 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. When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions or operational situations, is discovered not to have been performed when specified, SR 3.0.3 allows the full delay period of 24 hours to perform the Surveillance. SR 3.0.3 also provides a time limit for completion of Surveillances that become applicable as a consequence of'NODE changes imposed by Required Actions.Failure to comply with specified Frequencies for SRs is expected to be an'nfrequent 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. 0 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 (continued) BFN-UNIT 2 B 3.0-13 Amendment il~il SR Applicability B 3.0 BASES SR 3.0.3 (continued) 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 a1lowed'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. 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 d'oes 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 instance, LCO 3.0.4 will govern any restrictions that may (or may not)apply to MODE or other specified condition changes.The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability (continued) BFN-UNIT 2 B 3.0-14 Amendment il il 41 SR Applicability B 3.0 SR 3.0.4 (continued) 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.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 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'nnotation is found in Section 1.4, Frequency. SR 3.0..4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from NODE 2.Furthermore, SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3.The requirements of SR 3.0.4 do not apply in NODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3)because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.BFN-UNIT 2 B 3.0-15 Amendment ik SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)BASES BACKGROUND SDM requirements are specified to ensure: a~b.The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;'The reactivity transients associated with postulated accident conditions are controllable within.acceptable limits;and c.The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. These requirements are satisfied by the control rods, as described in GDC 26 (Ref.1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions. APPLICABLE.SAFETY ANALYSES The control rod drop accident (CRDA)analysis (Refs.2 and 3)assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO 3.1.6,"Rod Pattern Control").Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref.4)and fuel assembly insertion error during refueling (Ref.5 accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling.(Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO 3.10.6,"Multiple Control Rod Withdrawal -Refueling.") The analysis assumes this condition is acceptable since the core will be (continued) BFN-UNIT 2 B 3.1-1 Amendment 0 SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) shut down with the highest worth control rod withdrawn, if adequate SDM has been demonstrated. Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity. Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage.SDM satisfies Criterion 2 of the NRC Policy Statement (Ref.8).LCO The specified SDH limit accounts for the uncertainty in the demonstration of SDM by testing.Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDH test when the highest worth control rod is determined by measurement. When SDM is demonstrated by calculations not associated with a test (e.g., to confirm SDH during the fuel loading sequence), additional margin is included to account for uncertainties in the calculation. To ensure adequate SDM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref.6).APPLICABILITY In MODES 1 and 2, SDM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref.2).In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod.SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or a fuel assembly insertion error (Ref.5).(continued) BFN-UNIT 2 B 3.1-2 Amendment 0 0 SDM B 3.1.1 BASES (continued) ACTIONS A.l With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours.Failure to meet the specified SDM may be caused by a control rod that cannot be inserted.The allowed Completion Time of 6 hours is acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, and the low probability of an event occurring during this interval.B.l If the SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours, to prevent the potential for further reductions in available SDN (e.g., additional stuck control rods).The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach NODE 3 from full power conditions in an orderly manner and without challeng'ing plant systems.0 C.l With SDN not within limits in MODE 3', the operator must immediately initiate action to fully insert all insertable control rods.Action must continue until all insertable control rods are ful.ly inserted.This action results in the least reactive condition for the core.D.1 0.2 D.3 and D.4 With SDM not within limits in NODE 4, the operator must immediately initiate action to fully insert all insertable control rods.Action must continue until all insertable control rods are fully inserted.This action results in the.least reactive condition for the core.Action must also be initiated within 1 hour to provide means for control of potential radioactive releases.This includes ensuring secondary containment is OPERABLE;at least two Standby Gas Treatment (SGT)subsystems are OPERABLE;and secondary containment isolation capability (i.e., at least one secondary containment isolation valve (damper)and (continued).BFN-UNIT 2 B 3.1-3 Amendment ~~'k~ SDN B 3.1.1 BASES ACTIONS 0.1 0.'2 0.3 and 0.4 (continued) associated instrumentation are OPERABLE, or other acceptable administrative controls'to assure isolation capability) in each associated secondary containment penetration flow path with isolation valve(s)(damper(s)) not isolated that is assumed to be isolated to mitigate radioactive releases.This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons.It is not necessary to perform the surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component'is inoperable, then it must be restored to OPERABLE status.In this case, SRs may need to be performed to restore the component to OPERABLE status.Actions must continue until all required components are OPERABLE.E.1 E.2 E.3'E.4 and E.5 With SDH not within limits in NODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM (e.g., insertion of fuel in the core or the withdrawal of control rods).Suspension of these activities shall not preclude completion of movement of a component to a safe condition. Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions.Action must also be immediately initiated to fully inse}t all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted.Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted.Action must also be initiated within 1 hour to provide means for control of potential radioactive releases.This includes ensuring secondary containment is OPERABLE;at least two SGT subsystems are OPERABLE;and secondary containment isolation capability (i.e., at least one secondary containment isolation valve (damper)and associated instrumentation are OPERABLE, or other acceptable (continued) BFN-UNIT 2'8 3.1-4 Amendment ~i.ik Ib SDM B 3.1.1 BASES ACTIONS E.1 E.2 E.3 E.4 and E.5 (continued) administrative controls to assure isolation capability) in each associated secondary containment penetration flow path with isolation valve(s)(damper(s)) not isolated that is assumed to be isolated to mitigate radioactivity releases.This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons.It is not.necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status.In this case, SRs may need to be performed to restore the component to OPERABLE status.Action must continue until all required components are OPERABLE.SURVEILLANCE RE(UIREHENTS SR 3.1.1.1 Adequate SDH must be verified to ensure that the reactor can be made subcritical from any initial operating condition. This can be accomplished by a test, an evaluation, or a combination of the two.Adequate SDH is demonstrated before or during the first startup after fuel movement, or shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location.Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC)test must also account for changes in core reactivity during the cycle.Therefore, to obtain the SDH, the initial measured value must be increased by an adder,"R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required (Ref.7).For the SDH demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10%M/k)must be added to the SDH limit of 0.28%M/k to account for uncertainties in the calculation.(continued) BFN-UNIT 2 B 3.1-5 Amendment il~~i SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS SR 3.1.1.1 (continued) The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by testing.Local critical tests require the withdrawal of out of sequence control rods.This testing would therefore require bypassing of the rod worth minimizer to allow the out of sequence withdrawal, and therefore additional requi}ements must be met (see LCO 3.10.7,"Control Rod Testing-Operating").The Frequency of 4 hours after, reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification. During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel within the core)is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence.These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle.A spiral.reload sequence does not preclude the practice of bridging between SRMs and filling in the center in order to provide for conservative core monitoring during core alterations. Removing fuel from the core will always result in an increase in SDM.REFERENCES l.10 CFR 50, Appendix A, GDC 26.2.FSAR, Section 14.6.2.(continued) BFN-UNIT 2 B 3.1-6 Amendment il SDM B 3.1.1 BASES REFERENCES (continued) 3.NEDE-24011-P-A-11-US,"General Electric Standard Appl.ication, for Reactor Fuel," Supplement for United States, Section S.2.2.3.1, November 1995.4.FSAR, Section 14.5.3.3.5.FSAR, Section 14.5.3.4.6.FSAR, Section 3.6.5.2.7.NEDE-24011-P,-A-11,"General.Electric Standard Application for Reactor Fuel," Section 3.2.4.1, November 1995.8.NRC 93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.(continued) BFN-UNIT 2 B 3.1-7'mendment 0 ik Reactivity Anomalies B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEHS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with 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 abnormal operational transients. Therefore, reactivity anomaly is used as a measure of the predicted versus measured core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA)and transient safety analyses remain valid.A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or 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 SDH 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 SDH demonstrations (LCO 3.1.1,"SHUTDOWN HARGIN (SDH)")in assuring 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. In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC).When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable absorbers (e.g., gadolinia), control rods, and whatever neutron (continued) BFN-UNIT 2 B 3.1-8 Amendment 0!5 Reactivity Anomalies B 3.1.2 BASES BACKGROUND (continued) poisons (mainly xenon and samarium)are present in the fuel.The predi'cted core reactivity, as represented by control rod density, is calculated by a 3D core simulator code as a function of cycle exposure.This calculation is performed for projected operating states and conditions throughout the cycle.The core reactivity is determined from control rod densities for actual plant conditions and is then compared to the predicted value for the cycle exposure.APPLICABLE SAFETY ANALYSES Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations (Ref.2).In particular, SDM and reactivity transients, such as control.rod withdrawal accidents or rod drop 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 anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity. The comparison between measured and predicted initial cor e reactivity. provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted rod density for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used~to predict rod density 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 value.Thereafter, any significant deviations in the measured rod density from the predicted rod density that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.Reactivity anomalies satisfy Criterion 2 of the NRC Policy Statement (Ref.3).BFN-UNIT.2 B 3.1-9 (continued) Amendment il~il 0 Reactivity Anomalies B 3.1.2 LCO The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses.Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA'nd transient analyses are no longer valid, or that the uncertainties in the"Nuclear Design Methodology" are larger than expected.A limit on the difference between the monitored and the predicted rod density corresponding to a reactivity difference of i 1%M/k has been established based on engineering judgment.A>I/o deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated. APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved.Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly.In MODE 2, control rods are typically being withdrawn during a startup.In MODES 3 and.4, all control rods are fully inserted and therefore the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In MODE 5, fuel loading results in a continually changing core reactivity. SDM requirements (LCO 3.1.1)ensure that fuel movements are performed within the bounds of the safety analysis, and an SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, shuffling). The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, the reactivity anomaly LCO is not applicable during these conditions. ACTIONS A.1 Should an anomaly develop between actual and expected critical rod configuration,'the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly.This (continued) BEN-UNIT 2 B 3.1-10 Amendment il~0 II Reactivity Anomalies B 3.1.2 BASES ACTIONS A.1 (continued) evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Neasured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 72 hours 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.B.1 If the core reactivity cannot be restored to within the 1%Zdc/k limit, the plant must be brought to a,NODE in which the LCO does not apply.To achieve this status, the plant must be brought to at least NODE 3 within 12 hours.The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach NODE 3 from full power conditions. in an orderly manner and without challenging plant systems.SURVEILLANCE.REQUIRENENTS SR 3.1.2.1 Verifying the reactivity difference between the actual critical rod configuration and the expected configuration is within the'limits of the LCO provides added assurance that plant operation is maintained within the assumptions of the DBA and transient analyses.The core monitoring software calculates the k-effective for the critical rod configuration and reactor conditions. A comparison of this calculated k-effective at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a significant amount.This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled within the core, or control rods are replaced or shuffled.Control rod (continued). BFN-UNIT 2 B 3.1-11 Amendment 0 il Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE RE(U I REM ENTS SR 3.1.2.1 (continued) replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another.core location.Also, core reactivity changes during the cycle.The 24 hour interval after reaching equilibrium conditions following a.startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicted rod density can be made.For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes)at a 751'RTP have been obtained.The 1000 NWD/T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity. This comparison, requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results.Therefore, the comparison is only done when in NODE 1.REFERENCES 1.10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.2.FSAR, Chapter 14.6.3.NRC'No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT,2 B 3.1-12 Amendment Cl Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of the control rod drive (CRD)System, which is the primary reactivity control system for the reactor.In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including abnormal operational transients, that specified acceptable fuel design limits are not exceeded.In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.The CRD System is designed to satisfy the requirements of GDC 26, GDC 27, GDC 28,'and GDC 29 (Ref.1).The CRD System consists of 185 locking piston control rod drive mechanisms (CRDMs)and a hydraulic control unit for each drive mechanism. The locking piston type CROM is a double acting hydraulic piston, which uses condensate water as the operating fluid.Accumulators provide additional energy for scram.An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion. This Specification, along with LCO 3.1.4,"Control Rod Scram Times," and LCO 3.1.5,"Control Rod Scram Accumulators," ensure that the performance of the control rods in the event of a Design Basis Accident (DBA)or transient meets the assumptions used in the safety analyses of References 2, 3, and 4.APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in the evaluations involving control rods are presented in References 2, 3, and 4.The control rods provide the primary means for.rapid reactivity control{reactor scram), for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.(continued) BFN-UNIT 2 B 3.1-13 Amendment 0 0 0 Control Rod OPERABILITY B 3.1.3 APPLICABLE SAFETY ANALYSES (continued) The capability to insert the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.Since the SDM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA)can possibly occur.Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function.The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL)(see Bases for SL 2.1.1,"Reactor Core SLs" and LCO 3.2.2,"MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1%cladding plastic strain fuel design limit (see Bases for LCO 3.2.1,"AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.3,"LINEAR HEAT GENERATION RATE (LHGR)"), and the fuel damage limit (see Bases for LCO 3.1.6,"Rod Pattern Control")during reactivity insertion events.The negative reactivity insertion (scram)provided by the CRD System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CRDA.The Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6.discuss in more detail how the SLs are protected by the CRD System.Control rod OPERABILITY satisfies Criterion 3 of the NRC Policy Statement (Ref.6).LCO The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod'osition. Accumulator OPERABILITY is addressed by LCO 3.1.5.The associated scram accumulator status for a control rod only affects the scram insertion times;therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to (continued) BFN-UNIT 2 B 3.1-14 Amendment i~0 Control Rod OPERABILITY B 3.1.3 BASES LCO (continued), satisfy the intended reactivity control requirements, strict control over the number and, distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.APPLICABILITY In MODES I and 2, the, control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these HODES.In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.This provides adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5,"Control Rod OPERABILITY-Refueling." ACTIONS The ACTIONS Table is.modified by a Note indicating that a separate Condition entry is allowed for each.control rod.This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod.Complying with the Required.Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.A.l A.2 A.3 and A.4 A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure.With a fully inserted control rod stuck, no actions are requi'red as long as the control rod remains fully inserted.The Required Actions are modified by a Note, which allows the rod worth minimizer (RWH)to be bypassed if required to allow continued operation. LCO 3.3.2.1,"Control Rod Block Instrumentation," provides additional requirements when the RWH is bypassed to ensure compliance with the CRDA analysis.Wi,th one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met.Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if a)'he stuck control rod occupies a location adjacent to.(continued) BFN-UNIT 2 B 3.1-15 Amendment O.ik' Control Rod OPERABILITY 8 3.1.3 BASES ACTIONS A.1 A.2 A.3 and A.4 (continued) two"slow" control rods, b)the stuck control rod occupies a location adjacent to one"slow" control rod, and the one"slow" control rod is a1so adjacent to another"slow" control rod, or c)the stuck control rod, occupies a location adjacent to one"slow" control rod when there is another pair of"slow" control rods adjacent to one another.The description of"slow" control rod is provided in LCO 3.1.4,"Control Rod Scram Times." In addition, the associated control rod drive must be disa'rmed in 2 hours.Hydraulically disarming does not normally include isolation of the cooling water.The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner.The control rod must be isolated from both scram and normal insert and withdraw pressure.Isolating the control rod from scram prevents damage to the CROM.Monitoring of'he insertion capability.of each withdrawn control'od must also be performed within 24 hour s'from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP)of, the RWM.SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control rods.Testing.each withdrawn control rod ensures that a generic problem does not exist.This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." The Required, Action A.3 Completion Time only begins upon discovery that THERMAL POWER is greater than the actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6)and the RWM (LCO 3.3.2.1).The allowed Completion Time of 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the RWM provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.To allow, continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours.Should a DBA or transient require a shutdown, to (continued) BFN-UNIT 2 8 3.1-16 Amendment il 0 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.1 A.2 A.3 and A.4 (continued) preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required.Therefore, the original SDM demonstration may not be valid.The SDM must therefore be evaluated (by measurement or analysis)with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn. The allowed Completion Time of 72 hours to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.B 1 and B2 With two or more withdrawn control rods stuck, the stuck control rods must be isolated from scram pressure within 2 hours and the plant brought to MODE 3 within 12 hours.The control rod must be isolated from both scram and normal insert and withdraw pressure.Isolating the control rod from scram prevents damage to the CRDM.The allowed Completion Time is acceptable, considering the low probability of a CRDA occurring during this interval.The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required.Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert.The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.C.l and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted (continued) BFN-UNIT 2 B 3.1-17 Amendment

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) within 3 hours and disarmed (electrically or hydraulically) within 4 hours.Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.The control rod is disarmed (electrically or hydraulically) to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves while maintaining cooling water to the CRD.The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.l is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWN is bypassed to ensure compliance with the CRDA analysis.The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.D.l and D.2 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA.At~10%RTP, the generic banked position withdrawal sequence (BPWS)analysis (Ref.5)requires inoperable control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal.Therefore, if two or more inoperable control rods are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status.Condition D is modified by a Note indicating that the Condition is not applicable when)10%RTP, since the BPWS is not required to be followed under these conditions, as described in the Bases for LCO 3.1.6.The allowed Completion Time of 4 hours is acceptable, considering the low probability of a CRDA occurring.(continued) BFN-UNIT 2 B 3.1-18 Amendment il Control Rod OPERABILITY B 3.1.3 ACTIONS (continued) E.l If any Required Action and associated Completion Time of Condition A, C, or D are not met, or there are nine or more inoperable control rods, the plant must be brought to a NODE in which the LCO does not apply.To achieve this status, the plant must be brought to MODE 3 within 12 hours.This'ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram)of the control rods.Below 10%RTP, the generic banked position-withdrawal sequence (BPWS)analysis (Ref.5)allows a maximum of eight bypassed control rods.The number of control, rods permitted to be inoperable when operating above 10%RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.1.3.1 The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns.Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods.The 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.The control rod may then be returned to its original (continued) BFN-UNIT 2 B 3.1-19 Amendment O~il Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.1.3.2 and SR 3.1.3.3 (continued) position.This ensures the control rod is not stuck and is free to insert on a scram signal.These Surveillances are not required when THERMAL POWER is less than or equal to the.actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of banked position withdrawal sequence (BPWS)(LCO 3.1.6)and the RWM (LCO 3.3.2.1).The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods.Partially withdrawn control rods are tested at a 31 day Frequency, based on the potential power reduction required to allow the control rod movement and considering the large testing sample of SR 3.1.3.2.Furthermore, the 31 day Frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability must be made and appropriate action taken.SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is c 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1'.4.3, and SR 3.1.4.4.The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1,"Reactor Protection System (RPS)Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8,"Scram Discharge Volume (SDV)Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function.The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which.shows scram times do not significantly change over an operating cycle.(continued) BFN-UNIT 2 B 3.1-20 Amendment ll~il Control Rod OPERABILITY B'.1.3 SURVEILLANCE RE(UIREHENTS (continued) SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position.The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position.The verification is required to be performed any time a control rod is withdrawn to.the"full out" position (notch position 48)or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling.This includes control rods inserted one notch and then returned to the"full out" position during the'performance of SR 3.1.3.2.This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.REFERENCES 1.10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, and GDC 29.2.FSAR, Section 3.4.6.3.FSAR, Section 14.5.4.FSAR, Section 14.6.5.NED0-21231,"Banked Positi,on Withdrawal Sequence," Section 7.2, January.1977.6.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.1.-21 Amendment 0 0'0 Control Rod Scram Times B 3.1.4 B 3.1.4 Control Rod Scram Times BASES BACKGROUND The scram function of the Control Rod Drive (CRD)System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Ref.1).The control rods are ,scrammed by positive means using hydraulic pressure exerted on the CRD piston.When a scram signal is initiated, control air is vented from the scram valves, all'owing them to open by spring action.Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston.Since the notches in the index tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston.As the drive moves upward and the accumulator pressure reduces below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action.If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure.ES APPLICABLE SAFETY ANALYS The analytical methods and assumptions used in evaluating the control rod scram function are presented. in References 2, 3, and 4.The Design Basis Accident (DBA)and transient analyses assume that all of the control rods scram at a specified insertion rate.The resulting negative scram reactivity forms the.basis for the determination of plant thermal limits{e.g., the NCPR).Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time)can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the ,DBA and transient analyses can be met.{continued) BFN-UNIT 2 B 3.1-22 Amendment O~I~il Control Rod Scram Times~B 3.1.4 APPLICABLE SAFETY ANALYSES (continued) The scram function of the CRD System protects the MCPR Safety Limit (SL)(see Bases for SL 2.1.1,"Reactor Core SLs," and LCO 3.2.2,"MINIMUM CRITICAL POWER RATIO (MCPR)")and the 1%cladding plastic strain fuel design limit (see Bases for LCO 3.2.1,"AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded.Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref.5)and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6,"Rod Pattern Control").For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref.7).LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO)are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref.6).To account for single failures and"slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis.The scram times have a margin that allows up to approximately 7%of the control rods (e.g., 185 x 7%~13)to have scram times exceeding the specified limits (i.e.,"slow" control rods)assuming a single stuck control rod (as allowed by LCO 3.1.3,"Control Rod OPERABILITY")and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times.The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens (continued) BFN-UNIT 2 B 3.1-23 Amendment il Control Rod Scram Times B 3.1.4 BASES LCO (continued) ,("dropout") as the index tube travels upward.Verification of the specified scram times in Table 3.1.4-1, is accomplished through measurement of the"dropout" times.To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed"slow" control rods may occupy adjacent locations. Table 3.1.4-1 is modified by two Notes, which state that control rods with, scram times not within the limits of the table are considered"slow" and that control rods with scram times)7 seconds are considered inoperable as required by SR 3.1.3.4.This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3)'.Slow scramming control rods can be conservatively declared inoperable and not accounted for as"slow" control rods.APPLICABILITY In NODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these NODES.In NODES 3 and 4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod bl.ock is applied.This provides adequate requirements for control rod scram capability during these conditions. Scram requirements in NODE 5 are contained in LCO 3.9.5,"Control'Rod OPERABILITY-Refueling." ACTIONS A.1 When the requirements of this LCO are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analysis.Therefore, 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 NODE 3 within 12 hours.The allowed Completion Time of 12 hours is reasonable, based.on operating experience, to reach NODE 3 from full'ower conditions in an orderly manner and without challenging plant systems.BFN-UNIT 2 B 3.1-24 (continued) Amendment ik' Control Rod Scram Times 8 3.1.4 SURVEILLANCE REQUIREMENTS The four SRs of this LCO are modified by a Note stating that during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed)the influence, of the CRD pump head does not affect the single control rod scram times.During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time.Measurement of the scram times with reactor steam dome pressure a 800 psig demonstrates acceptable scram times for the transients analyzed in References 3 and 4.Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy.Therefore, demonstration of adequate scram times at reactor steam dome pressure w 800 psig ensures that the measured scram times will be, within the speci.fied limits at higher pressures. To ensure that scram time testing is performed within a reasonable time following fuel movement within the reactor pressure vessel after a shutdown~120 days or longer, control rods are required to be tested before exceeding 40%RTP following the shutdown.The SR is modified by a Note stating that in the event fuel movement is limited to selected core cells, only those CRDs associated with the core cells affected by the fuel movements are required to'be scram time tested.However, if the reactor remains shutdown a 120 days, all control rods are required to be scram time tested.This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control rods or the CRD System.(continued) BFN-UNIT 2 B 3.1-25 Amendment il 0 II Control Rod Scram Times'3.1.4 BASES SURVEILLANCE REQUIREMENTS SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle.A representative sample contains at least 10%of the control rods.This sample remains representative if no more than 20%of the control rods in the sample tested are determined to be"slow." With more than 20%of the sample declared to be"slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20%criterion (i.e., 20%of the entire sample)is satisfied, or until the total number of"slow" control rods (throughout the core from al'l Surveillances) exceeds the LCO limit.For planned testing, the control rods selected for the sample should be different for each test.Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample.The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle.This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more frequent intervals in-accordance with LCO 3.1.3 and LCO 3.1.5,"Control Rod Scram.Accumulators." SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure.The scram testing must be performed once before declaring the control rod OPERABLE.The required scram testing must demonstrate that for the affected control rod the scram valves open and the scram discharge path is open.This test can be performed with the control rod inserted and the accumulator drained and isolated to minimize potential damage to the drive.The test is adequate based on a high probability of meeting the scram.time testing acceptance criteria at reactor pressures e 800 psig.Limits for)800 psig are found in Table 3.1.4-1.(continued) BFN-UNIT 2 B 3.1-26 Amendment il~il~ Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE RE(UI RE>IENTS'SR 3.1.4.3 (continued) Specific examples of work that could affect the scram times are (but are'not limited to)the following: removal of any CRD for maintenance or modification; replacement of a control rod;and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the 1'imits of Table 3.1.4-1 with the reactor steam dome pressure a 800 psig.Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test.For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be required.This testing ensures.that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria.The Frequency of once prior to exceeding 40%RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. REFERENCES l.10 CFR 50, Appendix A, GDC 10.2.FSAR, Section 3.4.6.3.FSAR, Section 14.5.(continued) BFN-UNIT 2 B 3.1-27 Amendment /i il~il Control Rod Scram Times B 3.1'.4 BASES REFERENCES (continued) 4.FSAR, Section 14.6., 5.NEDE-'24011-P-A-11,"General Electric Standard Application for'Reactor Fuel," Section 3.2.4.1, November 1995.6..Letter from R.F.Janecek (BWROG)to R.W.Starostecki (NRC),"BWR Owners Group Revised Reactivity Control System, Technical Specifications,".BWROG-'8754, September 17, 1987.7..NRC No..93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.1-28 ,Amendment il~~i.ik Control Rod Scram Accumulators B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD)System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure.The accumulator is a hydraulic cylinder with a free floating piston.The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy.The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.4,"Control Rod Scram Times." APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the control rod scram function are presented in References 1, 2, and 3.The Design Basis Accident (DBA)and transient analyses assume that all of the control rods scram at a specified insertion rate.OPERABILITY of each individual'ontrol rod scram accumulator, along with LCO 3.1.3,"Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met.The existence of an inoperable accumulator may,invalidate prior scram time measurements for the associated control rod.The scram function of the CRD System, and therefore the OPERABILITY of the accumulators, protects the HCPR Safety Limit (see Bases for SL 2.1.1,"Reactor Core SLs" and LCO 3.2.2,"MINIMUM CRITICAL POMER RATIO (HCPR)")and 1%cladding plastic strain fuel design limit (see Bases for LCO 3.2.1,"AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.3,"LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4).In addition, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6,"Rod Pattern Control").(continued) BFN-UNIT 2 8 3.1-29 Amendment -45 0 ik Control Rod Scram Accumulators B 3.1.5 BASES APPLICABLE SAFETY ANALYSES (continued) Control rod scram accumulators satisfy Criterion 3 of the NRC Policy Statement (Ref.4).LCO The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.APPLICABILITY In MODES 1 and 2, the scram function is required for mitigation of DBAs and transients, and therefore the scram accumulators must be OPERABLE to support the scram function.In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is.applied.Requirements for scram accumulators in MODE 5 are contained in LCO 3.9.5,"Control Rod OPERAB I LITY-Re fuel i ng." ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod scram accumulator. This is acceptable since the Required Actions for each Condition provide appropriate compensatory actions for each affected accumulator. Complying with the Required Actions may allow for continued operation and subsequent affected accumulators governed by subsequent Condition entry and application of associated Required Actions.A.l and A.2 With one control rod scram accumulator inoperable and the reactor steam dome pressure a 900 psig, the control rod may be declared"slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the required scram times in Table 3.1.4-1.(continued) BFN-UNIT 2 B 3.1-30 Amendment I 0 Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS A.I and A.2 (continued) Required Action A.l is modi.fied by a Note indicating that declaring the control rod"slow" only applies if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time test.Otherwise, the control rod would already be considered"slow" and the further degradation of scram performance with an inoperable accumul.ator could result in excessive scram times.In this event, the associated control rod is declared inoperable (Required Action A.2)and LCO 3.1.3 is entered.This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function, in accordance with ACTIONS of LCO 3.1.3.The allowed Completion Time of 8 hours is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures. B.l B.2.1 and B.2.2 With two or more control rod scram accumulators inoperable and reactor steam dome pressure a 900 psig, adequate pressure must be supplied to the charging water header.With inadequate charging water pressure, all of the accumulators could become inoperable, resulting in'a potentially severe degradation of the scram performance. Therefore, within.20 minutes from discovery of charging water header pressure (.940 psig concurrent with Condition B, adequate charging water header pressure must be restored.The allowed Completion Time of 20 minutes is reasonable, to place a CRD pump into service to restore the charging water header pressure, if required.This Completion Time is based on the ability of the reactor pressure alone to fully insert all control rods.The control rod may be declared"slow," since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 3.1.4-1.Required Action B.2.1 is modified by a Note indicating that declaring the control rod"slow" only applies if the associated control scram time is within the limits of Table 3.1.4-1 during the last scram time test.Otherwise, the control rod (continued) BFN-UNIT 2 B 3.1-31 Amendment 'gi tJ 0 4l Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS B.1 B.2.1 and B.2.2 (continued) -would already be considered"slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times.In this event, the associated control rod is declared inoperable (Required Action B.2.2)and LCO 3.1.3 entered.This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3.The allowed Completion Time of 1 hour is reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable. C.l and C.2 With one or more control rod scram accumulators inoperable and the reactor steam dome pressure<900 psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged.With the reactor steam dome pressure<900 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during a depressurization event or at low reactor pressur es.Therefore, immediately upon discovery of charging water header pressure<940 psig, concurrent with Condition C, all control rods associated with inoperable accumulators must be verified tobe fully inserted.Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions. The associated control rods must also be declared inoperable within 1 hour.The allowed Completion Time of 1 hour is reasonable for Required Action C.2, considering the low probability of a DBA or transient occurring during the time that the accumulator is inoperable. D.l The reactor mode switch must be immediately placed in the shutdown position if either Required Action and associated Completion Time associated with the loss of the CRD charging pump (Required Actions B.-l and C.l)cannot be met.This (continued) BFN-UNIT 2 B 3.1-32 Amendment 0 Control Rod Scram Accumulators B 3.1.5 ACTIONS, D.l (continued) ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e., scram)of the control rods.This Required Action is modified by a Note stating that the action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed. SURVEILLANCE REQUIREMENTS SR 3.1.'5.1 SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force.An automatic accumulator monitor may be used to continuously satisfy this requirement. The primary indicator of accumulator OPERABILITY is the accumulator pressure..A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig (Ref.1).Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur.The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.REFERENCES 1.FSAR, Section 3.4.6.2.FSAR, Section 14.5.3.FSAR,, Section 14.6.4.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.1-33 Amendment il~gg~il Rod Pattern Control B 3.1.6 B'3.1 REACTIVITY CONTROL SYSTEHS B 3.1.6 ,Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWH)(LCO 3.3.2.1,"Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10%RTP.The sequences limit the potential amount o'f:reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1 and 2.'APPLICABLE 1 SAFETY ANALYSES The analytical methods and assumptions used in evaluating the CRDA are summarized in References 1 and 2.CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.The RWH (LCO 3.3.2.1)provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for UO~have been shown to be insignificant below fuel energy depositions of 300 cal/gm (Ref.3), the fuel damage limit of 280 cal/gm provides a-margin of safety from significant core damage which would result in release of radioactivity (Refs.4 and 5).Generic evaluations (Refs.1 and 6)of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm)have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASHE Code limits (Ref.7)and the calculated offsite doses will be wel.l within the required limits (Ref.5).(continued) 'BFN-UNIT 2'B 3.1-34 Amendment 0 il~0 Rod Pattern Control B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued) Control rod patterns analyzed in Reference 1 follow the banked position withdrawal sequence (BPWS).The BPWS is applicable from the condition of all control rods fully inserted to 10%RTP (Ref.2).For the BPWS, the control rods are.required to be moved in.groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12).The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. Generic analysis of the BPWS (Ref.8)has demonstrated that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generi'c BPWS analysis (Ref.8)allows a limited number (i.e., eight)and corresponding -distribution of fully inserted, inoperable control rods, that are not in compliance with the sequence.Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref.9).Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS.This LCO only applies to OPERABLE control rods.For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3,'"Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.'APPLICABILITY In MODES 1 and 2, when THERMAL POWER is x 10%RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required.When THERMAL POWER is)10%RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref.2).In, MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn. BFN-UNIT 2 B 3.1-35 (continued) Amendment ~i Oi 0 Rod Pattern Control B 3.1.6 BASES (continued) ACTIONS A.l and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence (Ref.8), actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours.Noncompliance with the prescribed sequence may be the result of"double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to a-10%RTP before establishing the correct control rod pattern.The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence.When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement must be stopped except for moves needed to correct the rod pattern, or scram if warranted. Required Action A.l is modified by a Note which allows the RWH to be bypassed to allow the affected control rods to be returned to their correct position.LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff.This ensures that the control rods will be moved to the correct position.A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2.The allowed Completion Time of 8 hours is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.B.l and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence (Ref.8).Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from, the prescribed sequence.Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of (continued) BFN-UNIT 2 B 3.1-36 Amendment ~~i il Rod Pattern Control B 3.1.6 ACTIONS B.1 and B.2 (continued) control rods has less impact on control rod worth than withdrawals have.Required Action B.1 is modified by a Note which al.lows the RWM to be bypassed to allow the affected control rods to be returned to their correct position.LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff.When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour.With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO.The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.SURVEILLANCE REgUIRENENTS SR 3.1.6.1 The control rod pattern is verified to be in compliance with the BPWS at a 24 hour Frequency to ensure the assumptions of the CRDA analyses are met.The 24 hour Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWH (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at a 10%RTP.REFERENCES 1.NEDE-24011-P-A-11-US,"General Electric Standard Application for Reactor Fuel, Supplement for United States," Section 2.2.3.1, November 1995.2.Letter from T.Pickens (BWROG)to G.C.Lainas (NRC), Amendment 17 to General Electric Licensing Topical Report, NEDE-24011-P-A, August 15, 1986.3.NUREG-0979, Section 4.2.1.3.2, April 1983.4.NUREG-0800, Section 15.4.9, Revision 2, July 1981.(continued) BFN-UNIT 2 B 3.1-37 Amendment il Rod'Pattern Control B 3.1.6 BASES REFERENCES (continued)'. 10 CFR 100.11.6.NED0-21778-A,"Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Mater Reactors," December 1978.7..ASNE, Boiler and Pressure Vessel Code.8.NED0-21231,"Banked Position Withdrawal Sequence," January 1977.9.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.1-38 Amendment 0'~0 SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC)System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement.The SLC System satisfies the requirements of 10 CFR 50.62 (Ref.1)on anticipated transient without scram.The SLC System consists of a boron solution storage tank, two positive displacement pumps in parallel and two explosive valves in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV).The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the'core.A smaller tank containing demineralized water is provided for testing purposes.The worst case sodium pentaborate solution concentration required to shutdown the reactor with sufficient margin to account for 0.05 Zb/k and Xenon poisoning effects is 9.2 weight percent.This corresponds to a 40'F saturation temperature. The worst case SLCS equipment area temperature is not predicted to fall below 50'F.This provides a 10'F thermal margin to unwanted precipitation of the sodium pentaborate. Tank heating components provide backup assurance that the sodium pentaborate solution temperature will never fall below 50'F but are not required for TS operability considerations. APPLICABLE SAFETY ANALYSES The SLC System is manually initiated from the main control room, as directed by the emergency operating instructions, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods.The SLC System is used in the event that enough control rods cannot be (continued) BFN-UNIT 2 B 3.1-39 Amendment il~II 0 SLC System B 3.1.7 BASES APPLICABLE SAFETY ANALYSES (continued) inserted to accomplish shutdown and cooldown in the normal manner.The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 660 ppm of natural boron, in the reactor coolant at 70'F.To allow for imperfect mixing, leakage and the volume in other piping connected to the reactor system, an amount of boron equal to 25%of the amount cited above is added (Ref.2).This volume versus concentration limit and the temperature versus concentration limits in Figure 3.1.7-1 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the water.volume in the entire residual heat removal shutdown cooling piping and in the recirculation loop piping.This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank.No credit is taken for the portion of the tank volume that cannot be injected.The SLC System satisfies Criterion 4 of the NRC Policy Statement (Ref.3).LCO The OPERABILITY of the SLC System provides backup capability.for reactivity control independent of normal reactivity control provisions provided by the control rods.The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the,RPV, including the OPERABILITY of the pumps and valves.Two SLC subsystems are required to be OPERABLE;each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.APPLICABILITY In MODES 1 and 2, shutdown capability is required.In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5, (continued) BFN-UNIT 2 B 3.1-40 Amendment II 0 0 SLC System B 3.1.7 APPLICABILITY (continued) only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1,"SHUTDOWN MARGIN (SDM)")ensures that the reactor will not become critical.Therefore, the SLC System is not required to be OPERABLE when on'ly a single control rod can be withdrawn. ACTIONS A.1 If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function.However, the overall reliability is reduced because a single failure in.the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis Accident (DBA)or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD)System to shut down the plant.B.l If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours.The allowed Completion Time of 8 hours is considered acceptable given the.low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.C.1 If any Required Action and associated Completion Time is not met, 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 12 hours.The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.BFN-UNIT 2 B 3.1-41 (continued) Amendment 0 0 4l SLC System B 3.1.7 BASES SURVEILLANCE RE(UIREHENTS SR 3.1.7.1 SR 3.1.7.1 is a 24 hour Surveillance verifying the volume of the borated solution in the storage tank, thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. This Surveillance ensures that the proper borated solution volume is maintained. The sodium pentaborate solution concentration requirements (c 9.2%by weight)and the required quantity of Boron-10 (a 186 lbs)establish the tank volume requirement. The 24 hour Frequency is based on operating experience that has shown there are relatively slow variations in the solution volume.SR 3.1.7.2 SR 3.1.7.2 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will occur if required.An automatic continuity monitor may be used to continuously satisfy this requirement. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed.The 31 day Frequency is based on operating experience and has demonstrated the reliability of the explosive charge continuity. SR 3'.1.7.3 and SR 3.1.7.5 SR 3.1.-7.3 requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank.The concentration is dependent upon the volume of water and quantity of boron in the storage tank.SR 3.1.7.5 requires verification that the SLC system conditions satisfy the following equation: C)Q E>=1.0 ('f3 WT%)(86 GPM)(19.8 ATOM%)C=sodium pentaborate solution weight percent concentration g=SLC system pump flow rate in gpm'E=Boron-10 atom percent enrichment in the sodium pentaborate solution (continued) BFN-UNIT 2 B 3.1-42 Amendment iS~il~il SLC System B 3.1.7 SURVEILLANCE REQUIREMENTS SR 3.1.7.3 and SR 3.1.7.5 (continued) To meet 10 CFR,50.62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution.The atom percentage of natural B-10 is 19.8%.This equivalency requirement is met when the equation given above is satisfied. The equation can be satisfied by adjusting the solution concentration, pump flow rate or Boron-10 enrichment. If the results of the equation are (1, the SLC System is no longer capable of shutting down the reactor with the margin described in Reference 2.However, the quantity of stored boron includes an additional margin (25%)beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water, leakage, and the volume in other piping connected to the reactor system.The sodium pentaborate solution (SPB)concentration is allowed to be>9.2 weight percent provided the concentration and temperature of the sodium pentaborate solution are verified to be within the limits of Figure 3.1.7-1.This ensures that unwanted precipitation of the sodium pentaborate does not occur.SR 3.1.7.3 and SR 3.1.7.5 must be performed every 31 days or within 24 hours, of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.The 31 day Frequency of these Surveillances is appropriate because of the relatively slow variation of boron concentration between surveillances. SR 3.1.7.3 must be performed within 8 hours of discovery that the concentration is>9.2 weight percent and every 12 hours thereafter until the concentration is verified to be z 9.2 weight percent.This Frequency is"appropriate under-these conditions taking into consideration the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.(continued) BFN-UNIT 2 B 3.1-43 Amendment O~il~0 SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.1.7.4 This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days.The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water.Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for SR 3.1.7.1.The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances. SR 3.1.7.6 Demonstrating that each SLC System pump develops a flow rate a 39 gpm at a discharge pressure a 1275 psig ensures that pump performance has not degraded during the fuel cycle.This minimum pump f1ow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay.This test confirms one point on the pump design curve and is indicative of overall performance. The 18 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance. SR 3.1.7.7 and SR 3.1.7.8 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the, firing of an explosive valve.The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfu11y fired.The pump and explosive valve tested'hould be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intervals. The Surveillance may be performed in separate steps to (continued) BFN-UNIT 2 B 3.1-44 Amendment il SLC System B 3.1.7 SURVEILLANCE REQUIREMENTS SR 3.1.7.7 and SR 3.1.7.8 (continued) prevent injecting boron into the RPV.An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV.The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint. Demonstrating that all piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution.An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the storage tank.The 18 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the piping or by other means.SR 3.1.7.9 The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water.Isotopic tests on these chemicals to verify the actual B-10 enrichment must be performed at least every 18 months and after addition of boron to the SLC tank in order to ensure that the proper B-10 atom percentage is being used and SR 3.1.7.5 will be met.The sodium pentaborate enrichment must be calculated within 24 hours and verified by analysis within 30 days.REFERENCES 1.10 CFR 50.62.2.FSAR, Section 3.8.4.3.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July.23, 1993.BFN-UNIT 2 B 3.1-45 Amendment ik~0' SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV)Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in.the SDV to ensure that sufficient volume is available at all times to allow a complete scram.During a scram, the SDV vent and drain valves close to contain reactor water.The SDV is a volume of header piping that connects to each hydraulic control unit (HCU)and drains into an instrument volume.There are two SDVs (headers)and two instrument volumes, each receiving, approximately one half of the control rod drive (CRD)discharges. Each instrument volume is connected to the radwaste system by a drain line containing two valves in series.Each header is connected to a common vent line with two valves in series for a total of four vent valves.The header piping is sized to receive and contain all the water discharged by the CRDs during a scram.The design and functions of the SDV are described in Reference l.APPLICABLE The Design Basis Accident and transient analyses assume all SAFETY ANALYSES of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to: a.Close during scram to l.imit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100 (Ref.3);and b.Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during"a-scram. Isolation of the SDV can also be accomplished by manual closure of the SDV valves.Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves.The offsite doses resulting from reactor coolant discharge from the SDV are significantly lower than the bounding doses resulting from a main steam line break outside the secondary containment (Ref.2)and are well within the limits of (continued) BFN-UNIT 2 B 3.1-46 Amendment il~il' SDV Vent and Drain Valves 8 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued) 10 CFR 100 (Ref.3).Adequate core cooling is by the integrated operation of the Emergency Core Cooling Systems (Ref.4).The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation to ensure that the SDV has, sufficient capacity to contain the reactor coolant discharge dur'ing a full core scram.To automatically ensure this capacity, a reactor scram (LCO 3.3.1.1,"Reactor Protection System (RPS)Instrumentation")is initiated if the SDV water level in the instrument volume exceeds a specified setpoint.The setpoint is chosen so that all control rods are inserted before the SDV has insufficient volume to accept a full scram.SDV vent and drain valves satisfy Criterion 3 of the NRC Policy Statement (Ref.5).LCO The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping.Since each vent and drain line is provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system.Additionally, the valves are required to open on scram reset to ensure that a path is available for the SDV piping to drain freely at other times.APPLICABILITY In MODES 1 and 2, scram may be required;therefore, the SDV vent and drain valves must be OPERABLE.In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.This provides adequate controls to ensure that only a single control rod can be withdrawn.- Also, during MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram.(continued) BFN-UNIT 2 B 3.1-47 Amendment 0 II 0 SDV Vent and Drain Valves B 3.1.8 ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each SDV vent and drain line.This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SDV line.Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions.A.l When one SDV vent or drain valve is inoperable in one or more lines, the valve must be restored to OPERABLE status within 7 days.The Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a scram occurring during the time the valve(s)are inoperable. The SDV is still isolable since the redundant valve in the affected line is OPERABLE.During these periods, the single failure criterion may not be preserved, and a higher risk exists to allow reactor water out of the primary system during a scram.B.1 If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram.When a line i's isolated, the potential for an inadvertent scram due to high SDV level is increased. Required Action B.1 is modified by a Note that allows periodic draining and venting of the SDV when a line is isolated.During these periods, the line may be unisolated under administrative control.This allows any accumulated water in the line to be drained, to preclude a reactor scram on SDV high level.This is acceptable since the administrative controls ensure the valve can be closed quickly, by a dedicated operator, if a scram occurs with the valve open.The 8 hour Completion Time to isolate the line is based on the low probability of a scram occurring while the line is not isolated and unlikelihood of significant CRD seal leakage.(continued) BFN-UNIT 2 B 3.1-48 Amendment i~ SDV Vent and Drain Valves B 3.1.8 BASES ACTIONS (continued) If any Required Action and associated Completion Time is not met, the plant must be brought to a NODE in which the LCO does not apply.To achieve this status, the plant must be.brought to at least NODE 3 within 12 hours.The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach NODE 3 from full power condi.tions in an orderly manner and without challenging plant systems.SURVE ILL'ANCE REQUIREMENTS SR 3.1.8.1 During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2)to al,low for drainage of the SDV piping.Verifying that each valve is in the open position ensures that'the SDV vent and drain valves, will perform their intended functions, during normal operation. This SR does not require.any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation, which ensure correct valve positions. SR 3.1.8.2 During, a scram, the SDV vent and drain valves should close.to contain the reactor water discharged, to the SDV piping.Cycling each valve through its complete range of motion (closed and open)ensures that the valve will function.properly during a scram.The 92 day Frequency is based on operating experience and takes into account the level of redundancy in the system design.(continued). BFN-UNIT,2 B 3.1-49 Amendment ~i 0 SDV Vent and Drain Valves B 3.1.8 BASES'SURVE IL'LANCE RE(UIREMENTS (continued) SR 3.1.8.3 SR 3.1.8.3 is an integrated test of the: SDV vent and drain valves to'verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified.The closure time of 60 seconds after receipt of a scram signal is acceptable based on the bounding analysis for release of reactor coolant outside containment (Ref.2).Similarly, after receipt of.a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified.The LOGIC.SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the: assumed safety function.The 18 month Frequency is based on the need to perform.this Surveillance under the conditions that apply during a plant outage and the potential for an.unplanned. transient if the Surveillance were performed with the reactor at power., Operating experience has shown these components usually pass the Surveillance when performed at the 18 month'requency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1.FSAR, Section 3.4.5.3.1. 2.FSAR, Section 14.6.5.3.10 CFR 100.4.FSAR, Section 6.5.5.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2-B 3.1-50 Amendment il~il~il APLHGR B 3.2.1 B 3.2 POMER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)BASES BACKGROUND The APLHGR is a measure of.the average LHGR of all the fuel rods in a fuel assembly at any axial location.Limits on the APLHGR are specified to ensure that the fuel design 1imits identified in Reference I are not exceeded during abnormal operational transients and that the peak cladding temperature (PCT)during the postulated design basis loss of coolant accident (LOCA)does not exceed the limits specified in 10 CFR 50.46..~APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the fuel design limits are presented in References I and 2.The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), abnormal operational transients,,and normal operation that determine the APLHGR limits are presented in References I, 2, 3, and 4.Fuel design evaluations are performed to demonstrate that the 1%limit on the fuel cladding plastic strain and other fuel design limits described in Reference I are not exceeded during abnormal operational transients for operation with LHGRs up to the operating limit LHGR.APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly.APLHGR limits are developed as a function of exposure and fuel bundle type.LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46.The analysis is performed using calculational models.that are consistent with the requirements of 10 CFR 50, Appendix K.A complete discussion of the analysis code is provided in Reference 5.The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor.A (continued) BFN-UNIT 2 B 3.2-1 Amendment 0 il l APLHGR B 3.2.1 BASES APPLICABLE conservative, multiplier is applied to the LHGR assumed in SAFETY ANALYSES the LOCA analysis to account for the uncerta'inty associated (continued) with the measurement of the APLHGR.The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref.6)..LCO The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and'ransient analyses.APPLICABILITY The APLHGR limits are primarily derived from fuel.design evaluations and LOCA and transient analyses that are as'sumed to occur at high power levels.Design calculations (Ref.4)and operating experience have shown that as power is reduced, the margin to the.required APLHGR limits, increases. This trend continues down to the power range of'5%to 15%RTP when entry into MODE 2 occurs.When in MODE 2, the intermediate.range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2.Therefore, at THERMAL POWER levels x 25%RTP, the reactor is operating with substantial margin to the APLHGR limits;thus, this LCO is not required.ACTIONS A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met.Therefore, prompt action should be taken to restore the APLHGR(s)to within the required l.imits such that the plant operates within analyzed conditions,and: within design limits of the fuel rods.The 2 hour Completion Time is sufficient to restore the APL'HGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification. BFN-UNIT 2 B.3.2-2 (continued) Amendment Il il 0 APLHGR B 3.2.1 BASES (continued) ACTIONS (continued) SURVEILLANCE RE(UI REM ENTS B.l If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply.To achieve this status, THERMAL POWER must be reduced to<25%RTP within 4 hours.The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to<25%RTP in an orderly manner and without challenging plant systems.SR 3.2.1.1 APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is w 25%RTP and then every 24 hours thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis.The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour al.lowance after THERMAL POWER a 25%RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.REFERENCES 1.NEDE-24011-P-A-11"General Electric Standard Application for Reactor Fuel," November 1995.2.FSAR, Chapter 3.3.FSAR, Chapter 14.4.FSAR, Appendix N.5.6.NEDC-32484P,"Browns Ferry Nuclear Plant Units 1, 2;and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 1, February 1996.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 8 3.2-3 Amendment J, 0 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIHITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (HCPR)BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power.The HCPR Safety Limit (SL)is set such that 99.9%of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2).The operating limit HCPR is established to ensure that no fuel damage results during abnormal operational transients. Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref.1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs.Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling)for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the abnormal operational transients to establish the operating limit HCPR are presented in References 2, 3, 4, and 5.To ensure that the HCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power.ratio (CPR).The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.The limiting transient yields the largest change in CPR (hCPR).When the largest DCPR is added to the HCPR SL, the required operating limit HCPR is obtained.(continued) BFN-UNIT 2 B 3.2-4 Amendment il'l MCPR 8 3.2.2 APPLICABLE SAFETY ANALYSES (continued) Flow dependent correction factor for MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref.6)to analyze slow flow runout transients. The flow dependent correction factor is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref.7).LCO The MCPR operating limits spec'ified in the COLR are the resul.t of'he Design Basis Accident (DBA)and transient analysis.APPL'I CAB IL IT Y The MCPR operating limits are primarily derived, from transient analyses that are assumed to occur at high power levels.Below 25%RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small.Surveillance of thermal limits below.25%RTP is unnecessary due to the large inherent, margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.Statistical analyses indicate that the BFN-UNIT 2 B 3.2-5 (continued) Amendment ~i il~Cl MCPR B 3.2.2 BASES APPLICABILITY (continued) nominal value of the initial HCPR expected at 25%RTP is)3.5.Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the HCPR requirements, and that margins increase, as power is reduced to 25%RTP.This trend is expected to continue to the 5%to 15%power range when entry into NODE 2 occurs.When in HODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any HCPR compliance concern.Therefore, at THERHAL POWER levels<25%RTP, the reactor is operating with substantial margin to the HCPR limits and this LCO is not required.ACTIONS A.1 If any HCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.Therefore, prompt action should be taken to restore the HCPR(s)to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the HCPR(s)to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the HCPR out of specification. 8.1 If the HCPR cannot be restored to within its required limits within the associated Completion Time, the plant must.be brought to a HODE or other specified condition in which the LCO does not apply.To achieve this status, THERHAL POWER must be reduced to<25%RTP within 4 hours.The allowed Completion Time is reasonable, based on operating experience, to reduce THERHAL POWER to<25%RTP in an orderly manner and without challenging plant systems.(continued) BFN-UNIT 2 B 3.2-6 Amendment i 9 0' MCPR B 3.2.2 BASES (continued) SURVEILLANCE RE(UIREMENTS SR 3.2.2.1 The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is a 25%RTP and then every 24 hours thereafter. It is compared to the specified limits in the COLR to ensure that the reactor-is operating within the assumptions of the safety analysis.The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER a 25%RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.SR 3.2.2.2 determines the value of r, which is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,"Control Rod Scram Times")and Option B (realistic scram times)analyses.The parameter r must be determined once within 72 hours after each set of scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle.The 72 hour Completion Time is acceptable due to the relatively minor changes'in r expected during the fuel cycle.REFERENCES 1.NUREG-0562,"Fuel Rod Failure As a Consequence of Departure from Nucleate Boiling or Dryout,"-June 1979.2.NEDE-24011-P-A-11,"General Electric Standard Application for Reactor Fuel," November 1995.3.FSAR, Chapter 3.4.FSAR, Chapter 14.(continued) BFN-UNIT 2 B 3.2-7 Amendment II 0 MCPR B 3.2.2 BASES REFERENCES (continued) 5.FSAR, Appendix N.6.NED0-30130-A,"Steady State Nuclear Methods," May 1985..7.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.2-8 Amendment 0 0 LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location.Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including abnormal operational transients. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the anticipated operating conditions identified in Reference l.APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the fuel system design are presented in References 1 and 2.The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system)that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, 50, and 100.The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are: a.Rupture of the fuel rod cladding caused by strain from the relative expansion of the UO, pellet;and b;Severe overheating of the fuel rod cladding caused by inadequate cooling.A value of 1%plastic strain of the fuel cladding has been.defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref.3).Fuel design evaluations, have been performed and demonstrate that the 1%fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the (continued) BFN-UNIT 2 B 3.2-9 Amendment ll II LHGR B 3.2.3 BASES APPLICABLE SAFETY ANALYSES (continued) operating limit specified in the COLR.The analysis also'ncludes allowances for short term transient operation above the operating limit to account for abnormal operational transients, plus an allowance for densification power spiking.The'LHGR satisfies Criterion 2 of the NRC Policy Statement (Ref.4).LCO The LHGR is a basic assumption in the fuel design analysis.The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1%fuel cladding plastic strain.The operating limit to accomplish this objective is specified in the COLR.APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels<25%RTP,'the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at a 25%RTP.ACTIONS A.l If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met.Therefore, prompt action should:be taken to restore the LHGR(s)to within its required limits such that the plant is operating within analyzed conditions. The 2 hour Completion Time is.normally sufficient to restore the LHGR(s)to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification. B.1 If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the (continued) BFN-UNIT 2 B 3.2-10 Amendment il 0 Cl LHGR B 3.2.3 ACTIONS B.1 (continued) LCO does not apply.To,achieve this status, THERMAL POWER is reduced to<25%RTP within.4 hours.The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER TO<25%RTP in, an orderly manner and without.challenging plant systems.SURVEILLANCE RE(UIREMENTS SR 3.2.3.l The LHGR is required to'be initially calculated within 12 hours after THERMAL POWER is a 25%RTP and then every 24 hours thereafter. It is compared,to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis.The 24 hour Frequency is based on both engineering judgment and recognition of the: slow changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER a,25%RTP is achieved is acceptable given the large inherent'margin to operating limits at lower power levels.REFERENCES 1.FSAR,.Chapter 14..2.FSAR, Chapter 3.3.NUREG-0800, Standard Review Plan 4.2, Section II.A.2(g), Revision 2, July 1981.4.,NRC No.93-102,,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.2-11 Amendment 0 APRM Gain and Setpoints B 3.2.4 B 3.2 POWER DISTRIBUTION L'IHITS B 3.2.4 Average Power Range Monitor (APRH)Gain and Setpoints BASES BACKGROUND The OPERABILITY of the APRMs and their setpoints is an initial condition of all safety analyses that assume rod insertion upon reactor scram.Applicable GDCs are GDC 10,"Reactor Design," GDC 13,"Instrumentaf.ion and Control," GDC 20,"Protection System Functions," and GDC 23,"Protection System Failure Modes" (Ref.1).This LCO is provided to require the APRH gain or APRH flow biased scram setpoints to be adjusted when operating under conditions of excessive power peaking to maintain acceptable margin to the fuel cladding integrity Safety Limit (SL)and the fuel cladding 1%plastic strain limit.The condition of excessive power peaking is determined by the ratio of the actual power peaking to the limiting power peaking at RTP.This ratio is equal to the ratio of the core limiting HFLPD to the Fraction of RTP (FRTP), where FRTP is the measured THERMAL POWER divided by the RTP.Excessive power peaking exists when:)1, NFLPD FRTP indicating that HFLPD is not decreasing proportionately to the overall power reduction, or conversely, that power peaking is increasing. To maintain margins similar to those at RTP conditions, the excessive power peaking is compensated by a gain adjustment on the APRHs or adjustment of the APRM setpoints. Either of these adjustments has effectively the same result as maintai'ning HFLPD less than or equal to FRTP and thus maintains RTP margins for APLHGR and HCPR.The normally selected APRM setpoints position the, scram above the upper bound of the normal power/flow operating region that has been considered in the design of the fuel rods.The setpoints are flow biased with a slope that approximates the upper flow control line, such that an approximately constant margin is maintained between the flow biased trip level and the upper operating boundary for core flows in excess of about 45%of rated core flow.In the range of infrequent operations below 45%of rated core flow, (continued) BFN-UNIT 2 B 3.2-12 Amendment il~~~45 APRM Gain and Setpoints B 3.2.4 BASES BACKGROUND (continued) the margin to scram is reduced.because of the nonlinear core flow versus drive flow relationship. The normally selected APRM setpoints are supported by the analyses presented in References 1 and 2 that concentrate on events initiated from rated conditions. Design experience.has shown that minimum deviations occur within expected margins to operating limits (APLHGR and MCPR), at rated conditions for normal power distributions. However, at other than rated conditions, control rod patterns can be established that significantly reduce the margin to thermal limits.Therefore, the flow biased APRH scram setpoints may be reduced during operation when the combination of THERMAL POWER and HFLPD indicates an excessive power peaking distribution. The APRH neutron flux signal is also adjusted to more closely follow the fuel cladding heat flux during power transients. The APRH neutron flux signal is a measure of the core thermal power during steady state operation. During power transients, the APRM signal leads the actual core thermal power response because of the fuel thermal time constant.Therefore, on power increase transients, the APRH signal provides a conservatively high measure of core thermal power.By passing the APRM signal through an electronic filter with a time constant less than, but approximately equal to, that of the fuel thermal time constant, an APRM transient response that more closely follows actual fuel cladding heat flux is obtained, while a conservative margin is maintained. The delayed response of the filtered APRM signal allows the flow biased APRM scram level's to be positioned closer to the upper bound of the normal power and flow range, without unnecessarily causing reactor scrams during short duration neutron flux spikes.These spikes can be caused by insignificant transients such as performance of main steam line valve surveillances or momentary flow increases of only several percent.APPLICABLE SAFETY ANALYSES The acceptance criteria for the APRM gain or setpoint adjustments are that acceptable margins (to APLHGR and HCPR)be maintained to the fuel cladding integrity SL and the fuel cladding 1%plastic strain limit.FSAR safety analyses (Refs.2 and 3)concentrate on the rated power condition for.which the minimum expected margin to the operating limits (APLHGR and MCPR)occurs.(continued) BFN-UNIT 2 B 3.2-13 Amendment 0 0 0 APRH Gain and Setpoints B 3.2.4 APPLICABLE SAFETY ANALYSES (continued) LCO 3.2.1,"AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.2,"MINIMUM CRITICAL POWER RATIO (MCPR)," limit the initial margins to these operating limits at rated conditions so that specified acceptable fuel design limits are met during transients initiated from rated conditions. At initial power level's less than rated levels, the margin degradation of either the APLHGR or the HCPR during a transient can be greater than at the rated condition event.This greater margin degradation during the transient is primarily offset by the larger initial margin to limits at the lower than rated power levels.However, power distributions can be hypothesized that would result in reduced margins to the pre-transient operating limit.When combined with the increased severity of certain transients at other than rated conditions, the SLs could be approached. At substantially reduced power.levels, highly peaked power distributions could be obtained that could reduce thermal margins to the minimum levels required for transient events.To prevent or mitigate such situations, either the APRH gain is adjusted upward by the ratio of the core limiting HFLPD to the FRTP,, or the flow biased APRH scram level is required to be reduced by the ratio of FRTP to the core limiting MFLPD.Either of these adjustments effectively counters the increased severity of some events at other than rated conditions by proportionally increasing the APRH gain or proportionally lowering the flow biased APRH scram setpoints, dependent on the increased peaking that may be encountered. The APRH gain and setpoints satisfy Criteria 2 and 3 of the NRC Policy Statement (Ref.4).LCO Meeting any one of the following conditions ensures acceptable operating margins for events described above: a.Limiting excess power peaking;b.Reducing the APRM flow biased neutron flux upscale scram setpoints by multiplying the APRH setpoints by the ratio of FRTP and the core limiting value of HFLPD;or (continued) BFN-UNIT 2 B 3.2-14 Amendment 0 0 APRM Gain and Setpoints B 3.2.4 BASES LCO (continued) c.Increasing APRH gains to cause the APRH to read a 100 times HFLPD (in%).This condition is to account for the reduction in margin to the fuel cladding integrity SL and the fuel cladding 1%plastic strain limit.HFLPD is the ratio of the limiting LHGR to the LHGR limit for the specific bundle type.As power is reduced, if the design power distribution is maintained, HFLPD is reduced in proportion to the reduction in power.However., if power peaking increases above the design value, the HFLPD is not reduced in proportion to the reduction in power.Under these conditions, the APRM gain is adjusted upward or the APRH flow biased scram setpoints are reduced accordingly. When the reactor is operating with peaking less than the design value, it is not necessary to modify the APRM flow biased scram setpoints. Adjusting APRM gain or setpoints is equivalent to HFLPD less than or equal to FRTP, as stated in the LCO.For compliance with LCO Item b (APRH setpoint adjustment) or Item c (APRM'gain adjustment), only APRHs required to be OPERABLE per LCO 3.3.1.1,"Reactor Protection System (RPS)Instrumentation," are required to be adjusted.In addition, each APRH may be allowed to have its gain or setpoints adjusted independently of other APRHs that are having their gain or setpoints adjusted.APPLICABILITY The HFLPD limit, APRH gain adjustment, and APRH flow biased scram and associated setdowns are provided to ensure that the fuel cladding integrity SL and the fuel cladding 1%plastic strain limit are not violated during design basis transients. As discussed in the Bases for LCO 3.2.1 and LCO 3.2.2, sufficient margin to these limits exists below 25%RTP and, therefore, these requirements-are only necessary when the reactor is operating at a 25%RTP.ACTIONS A.l If the APRH gain or setpoints are not within limits while the HFLPD has exceeded FRTP, the margin to the fuel cladding integrity SL and the fuel cladding 1%plastic strain l,imit (continued) BFN-UNIT 2 B 3.2-15 Amendment 0 II APRM Gain and Setpoints B 3.2.4 BASES ACTIONS A.1 (continued) may be reduced.Therefore, prompt action should'e taken to restore the MFLPD to within its required limit or make acceptable APRH adjustments such that the plant is operating within the assumed margin of the safety analyses.The 6 hour Completion Time is normally sufficient to restore either the MFLPD to within limits or the APRH gain or setpoints to within limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LCO not met.B.l If MFLPD cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not-apply.To achieve this status, THERMAL POWER is reduced to<25%RTP within 4 hours.The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to<25%RTP in an orderly manner and without challenging plant systems.SURVEILLANCE RE(UIREHENTS SR 3.2.4.1 and SR 3.2.4.2 The HFLPD is required to be calculated and compared with FRTP, or APRH gains or setpoint, to ensure that the reactor is operating within the assumptions of the safety analysis.These SRs are only required to determine the,HFLPD and, assuming HFLPD is greater than FRTP, the appropriate gain or setpoint, and are not intended to be a CHANNEL FUNCTIONAL TEST for the APRH gain or flow biased neutron flux scram circuitry. The 24 hour Frequency o'f SR 3.2.4.1 is chosen to coincide with the determination of other thermal limits, specifically those for the APLHGR (LCO 3.2.1).The 24 hour Frequency is based on both engineering judgment and recognition of.the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER a 25%RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.BFN-UNIT 2 8 3.2-16 (continued) Amendment II I~0 APRH Gain and Setpoints B 3.2.4 BASES SURVEIL'LANCE REg UIREHENTS SR 3.2.4.1 and SR 3.2.4.2 (continued) The 12 hour'Frequency of SR 3.2.4.2 requires a more frequent verification than if HFLPD is less than or equal to FRP.When HFLPD is greater than FRP, more rapid changes in power distribution, are typically expected.REFERENCES 1.10 CFR 50, Appendix A, GDC 10, GDC 13, GDC 20, and GDC 23.2.FSAR, Chapter 14.3.FSAR, Chapter 3.4.NRC No.,93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.2-17 Amendment ik ib RPS Instrumentation 8 3.3.1.1 B 3.3, INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS)Instrumentation BASES BACKGROUND The RPS initiates a, reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS).and minimize the energy that must be absorbed following a loss of coolant accident (LOCA).This can be accomplished either automatically or manually.The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor.This is achieved by specifying limiting safety system settings (LSSS)in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system-parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs)during Design Basis Accidents (DBAs).The RPS, as described in the FSAR, Section 7.2 (Ref.1), includes sensors, relays, bypass circuits, and switches that are necessary to cause initiation of a reactor scram.Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam l,ine isolation valve position, turbine control valve (TCV)fast closure (indicated by TCV low hydraulic pressure)trip oil pressure, turbine stop valve (TSV)position, drywell pressure, scram pilot air header pressure, and scram discharge volume (SDV)water level, as well as reactor mode switch in shutdown position, manual, and RPS channel test switch scram signals.There are at least four redundant sensor input signals from each of these.parameters (with the exception of the reactor mode switch in shutdown and manual scram signals).Host channels include electronic equipment (e.g., trip units)that compares measured input signals with pre-established setpoints.. When the setpoint is exceeded, the channel output relay deenergizes actuates, which then outputs an RPS trip signal to the trip logic.(continued) BFN-UNIT 2 B 3.3-1 Amendment 0 RPS Instrumentation B 3.3.1.1 BASES BACKGROUND (continued) The RPS is comprised of two independent trip systems (A and B)with two logic channels in each trip system (logic channels Al and A2, Bl and B2)as shown in Reference 1.The outputs of the logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system.The tripping of both trip systems will produce a reactor scram.This logic arrangement is referred to as a one-out-of-two taken twice logic.Each trip system can be reset by use of a reset switch.If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for 10 seconds after the full scram signal is received.This 10 second delay on reset ensures that the scram function will be completed. Two scram pilot valves are located in the hydraulic control unit for each control rod drive (CRD).Each scram.pilot valve is solenoid operated, with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD.When either scram pilot valve solenoid is energized, air pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de-energized to cause a control rod to scram.The scram valves control the supply and discharge paths for the CRD water during a scram.One of the scram pilot valve solenoids for each CRD is controlled by trip system A, and the other solenoid is controlled by trip system B.Any trip of trip system A in conjunction with any trip in trip system B results in de-energizing both solenoids, air bleeding.off, scram valves opening, and control rod scram.The backup scram valves, which energize on a full scram signal to depressurize the scram air header, are also controlled by the RPS.Additionally, the RPS System controls the SDV vent and drain valves such that when both trip systems trip, the SDV vent and drain valves close to isolate the SDV.APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The actions of the RPS are assumed in the safety analyses of References 1, 2, and 3.The RPS initiates a reactor scram when monitored parameter values exceed the Allowable Values, specified by the setpoint methodology and listed in Table 3.3.1.1-1 to preserve the integrity of the fuel (continued) BFN-UNIT 2 B 3.3-2 Amendment 0 RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) cladding, the reactor coolant pressure boundary (RCPB), and the containment by minimizing the energy that must be absorbed following a LOCA.RPS instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.10).Functions not specifically credited in the accident analysis are retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.The OPERABILITY of the RPS is dependent on the-OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1. Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Allowable Values are specified for each RPS Function specified in the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibratio'n, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe (continued) BFN-UNIT 2 B 3.3-3 Amendment 0 0 RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.The individual Functions are required to be OPERABLE in the NODES or other specified conditions in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each NODE to provide primary and diverse initiation signals.The only MODES specified in Table 3.3.1.1-1 are NODES 1 (which encompasses >30%%u RTP)and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1)does not allow any control rod to be withdrawn. In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.Provided all other control rods remain inserted, no RPS function is required.In this condition, the required SDM (LCO 3.1.1)and refuel position one-rod-out interlock (LCO 3.9.2)ensure that no event requiring RPS will occur.The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.Intermediate Ran e Monitor IRH l., Intermediate an e Mo ito Neutron Fl x-Hi h The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRH)to the lower range of the average power range monitors (APRHs).The IRHs are capable of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients in the intermediate power range.In this power range, the most (continued) BFN-UNIT 2 B 3.3-4 Amendment il il RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSE LCO, and APPLICABILITY S, ntermed te a e o to Neut o x-(continued) significant source of reactivity change is due to control rod withdrawal. The IRM mitigates control rod withdrawal error events and is diverse from the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power.The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref.2).The IRM provides mitigation of the neutron flux excursion. To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref.3)to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the IRM.This analysis, which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against local control rod withdrawal errors and results in peak fuel energy depositions below the 170 cal/gm fuel failure threshold criterion. The IRMs are also capable of limiting other reactivity excursions during startup, such as cold water injection events;although no credit is specifically assumed.The IRM.System is divided into two groups of IRM channels, with four IRM channels inputting to each trip system.The analysis of Reference 3 assumes that one channel in each trip system is bypassed.Therefore, six channels with three channels in each trip system are required for IRM OPERABILITY. to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.This trip is active in each of the 10 ranges of the IRM, which must be selected by the operator to maintain the neutron flux within the monitored level of an IRM range.The analysis of Reference 3 has adequate conservatism to permit an IRM Allowable Value of 120 divisions of a 125 division scale.The Intermediate Range Monitor Neutron Flux-High Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists.In HODE 5, when a cell with fuel has its control rod withdrawn, the IRMs provide monitoring for and protection against (continued) BFN-UNIT 2 B 3.3-5 Amendment 0 ik RPS Instrumentation 8 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY I termediate Ran e Mo itor eutro ux-i (continued) unexpected reactivity excursions. In NODE 1, the APRN System and the RBM provide protection against control rod withdrawal error events and the IRMs are not required.b t ed te Ran e Mo to-o This trip signal provides assurance that a minimum number of IRNs are OPERABLE.Anytime an IRN mode switch is moved to any position other than"Operate," the detector voltage drops below a preset level, or when a module is not plugged in, an inoperative trip signal will be received by the RPS unless the IRN is bypassed.Since only one IRM in each trip system may be bypassed, only one IRN in eqch RPS trip system may be inoperable without resulting in an RPS trip signal.This Function was riot specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.Six channels of Intermediate Range Monitor-Inop with three channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.Since this Function is not assumed.in the safety analysis, there is no Allowable Value for this Function.This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux-High Function is required.Avera e Power Ran e Monitor 2.a.vera e Power Ran e o itor Neutron F ux-Hi h Setdown The APRN channels receive input signals from the local power range monitors (LPRHs)within the reactor core to provide an indication of the power distribution and local power changes.The APRN channels average these LPRN signals to provide a continuous indication of average reactor power from a few percent to greater than RTP.For operation at (continued) BFN-UNIT 2 B 3.3-6 Amendment I, I J il 0 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY vera e Powe Ran e o ito e t o u-h~Setdo (continued) low power (i.e., NODE 2), the Average Power Range Monitor Neutron Flux-High, Setdown Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range.For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High, Setdown Function will provide a secondary scram to the.Intermediate Range Monitor Neutron Flux-High Function because of the relative setpoints. With the IRNs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron Flux-High, Setdown Function will provide the primary trip signal for a corewide increase in power.No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux-High, Setdown Function.However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25/.RTP (SL 2.1.1.1)when operating at low reactor pressure and low core flow.Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER<25K RTP.The APRN System is divided into two groups of channels with three APRN channel inputs to each trip system.The system is designed to allow one channel in each trip system to be bypassed.Any one APRN channel in a trip system can cause the associated trip system to trip.Four channels of Average Power Range Monitor Neutron Flux-High, Setdown with two channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal.In addition, to provide adequate coverage of the entire core, at least 14 LPRN inputs are required for each APRN channel, with at least two LPRN inputs from each of the four axial levels at which the LPRNs are located.The Allowable Value is based on preventing significant increases in power when THERMAL POWER is<25%RTP.The Average Power Range Monitor Neutron Flux-High, Setdown Function must be OPERABLE during NODE 2 when control rods may be withdrawn since the potential for criticality exists.(continued) BFN-UNIT 2 B 3.3-7 Amendment 0'~', 0 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY t o F x-i.b.Avera e Power an e onitor Flow Biased S'ated T ermal Powe-Hi h.a vera e'Power a e Mon tor Setdown (continued) In MODE 1, the Average Power Range Monitor Neutron Flux-High Function provides protection against reactivity transients and the RWM and rod, block monitor, protect against control rod withdrawal error events.The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant.The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor.The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern)but is clamped at an upper limit that is always lower than or equal to the Average Power Range Monitor Fixed Neutron Flux-High Function Allowable Value.The'Average Power Range'Monitor Flow Biased Simulated Thermal Power-High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event)and protects the fuel cladding integrity by ensuring that the MCPR SL is not exceeded.During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram.For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Fixed Neutron Flux-High Function will provide a scram signal before the Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function setpoint is exceeded.The APRM System is divided into two groups of channels with three APRM channel inputs to each trip system.The system is designed to allow one channel in each trip system to be bypassed.Any one APRM channel in a trip system can cause the associated trip system to trip.Four channels of (continued) BFN-UNIT 2 B 3.3-8 Amendment ik 0 RPS Instrumentation B 3.3.l.l APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 2.b.Avera e Power Ran e Monitor Flow Biased Simulated Thermal Power-Hicih (continued) Average Power Range Monitor Flow Biased Simulated Thermal Power-High with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.Each APRM channel receives a total drive flow signal representative of total core flow.The total drive flow signals are generated by two flow units, one of which supplies signals to the trip system A APRMs, while the other one supplies signals to the trip system B APRMs.Each flow unit signal is provided by summing up the flow signals from the two recirculation loops.Each required Average Power Range Monitor Flow Biased Simulated Thermal Power-High channel requires an input from its associated OPERABLE flow unit.The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function for the mitigation of the loss of feedwater heating event.The THERMAL POWER time constant of<7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.The term"W" in the equation for determining the Allowable Value is defined as total recirculation flow in percent of rated.The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function is required to be OPERABLE in MODE I when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL).During NODES 2 and 5, other IRM and APRN Functions provide protection for fuel cladding integrity. 2.c.Avera e Power Ran e Monitor Fixed Neutron Flux-~Hi h The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The Average Power Range Monitor Fixed Neutron Flux-High Function is capable of generating a (continued) BFN-UNIT 2 B 3.3-9 Amendment 0 0 0 RPS Instrumentation B'.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY n e onitor ixed eutron F u-The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system.The system is designed to allow one channel in each trip system to be bypassed.Any one APRM channel in a trip system can cause the associated trip system to trip.Four channels of Average Power Range Monitor Fixed Neutron Flux-High with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.2,c Avera e Powe (continued) trip signal to prevent fuel damage or excessive RCS pressure.For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux-High Function is assumed to terminate the main steam isolation valve (MSIV)closure event and, along with the safety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV)pressure to less than the ASME Code limits.The control rod drop accident (CRDA)analysis (Ref.5)takes credit for the Average Power Range Monitor Fixed Neutron Flux-High Function to terminate the CRDA.The Allowable Value is based on the.Analytical Limit assumed in the CRDA analyses.The Average Power Range Monitor Fixed Neutron Flux-High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure)being exceeded.Although the Average Power Range Monitor Fixed Neutron Flux-High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-High, Setdown Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux-High Function is not required in MODE 2.(continued) BFN-UNIT 2 B 3.3-10 Amendment il RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY ,(continued) d ve a e Power Ran e Mo ito-Oow scale This signal ensures that there is adequate Neutron Monitoring System protection if the reactor mode switch is placed in the run position prior to the APRHs coming on scale.With the reactor mode switch in run, an APRM downscale signal coincident with an associated Intermediate Range Monitor Neutron Flux-High or Inop signal generates a trip signal.This Function was not specifically credited in.the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.The APRM System is divided into two groups of channels with three inputs into each trip system.The system is designed to allow one channel in each trip system to be bypassed.Four channels of Average Power Range Monitor-Downscale with two channels in each trip system arranged in a one-out-of-two logic are'.required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal.The Intermediate Range Monitor Neutron Flux-High and Inop Functions are also part of the OPERABILITY of the Average Power Range Monitor-Downscale 'Function (i.e., if either of these IRN Functions cannot send a signal to the Average Power Range Monitor-Downscale Function, the associated Average Power Range Monitor-Downscale channel is.considered inoperable). The Allowable Value is based upon ensuring that the APRNs are in the linear scale range when transfers are made between APRNs and IRNs.This Function is required to be OPERABLE in NODE 1 since this is when the APRNs are the primary indicators of reactor power.2.e.vera e Power Ran e Monitor-Ino This signal provides assurance that a minimum number of APRMs are OPERABLE.Anytime an APRH mode switch is moved to any position other than"Operate," an APRN module is unplugged, the electronic operating voltage is low, or the APRH has too few LPRH inputs (<14), an inoperative trip signal will be received by the RPS,'nless the APRN is bypassed.Since only one APRN in each trip system may be bypassed, only one APRN in each trip system may be (continued) BFN-UNIT 2 B 3.3-11 Amendment il RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY an e Monitor-no (continued) Four channels of Average Power Range Monitor-Inop with two channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal.There is no Allowable Value for this Function.2.e;Avera e Powe inoperable without resulting in an RPS trip signal.This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.This Function is required to be OPERABLE in the NODES where the APRM Functions are required.3 eactor Vesse Steam Dome Press e-i An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB.The Reactor Vessel Steam Dome Pressure-High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power.For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed'eutron Flux-High signal, not the Reactor Vessel Steam Dome Pressure-High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASHE Section III Code limits.High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure.The Reactor Vessel Steam Dome Pressure-High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.Four channels of Reactor Vessel Steam Dome Pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ('continued) BFN-UNIT 2 B 3.3-12 Amendment 0 RPS Instrumentation 8 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.Reactor Vessel Steam Dome Pressure-hicih (continued) ensure that no single instrument fai1ure will preclude a scram from this Function on a va1id signal.The Function is required to be OPERABLE in NODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.4.Reactor Vessel Water Level-Low Level 3 Low RPV water level indicates the, capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission.The Reactor Vessel Water Level-Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref.6).The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.Reactor Vessel Water Level-Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level, (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.The Reactor Vessel Water Level-Low, Level 3 Allowable Value is selected to ensure that (a)during normal operation the steam dryer skirt is not uncovered (this protects available recirculation pump net positive suction head (NPSH)from significant carryunder), and (b)for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water-Low Low Low, Level 1 will not be required.The.Function is required, in NODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor (continued) BFN-UNIT 2 B 3.3-13 Amendment !I i~II RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY actor esse Wate L'eve-ow eve 3 (continued) Vessel Water Level-Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other NODES.5.ai Stea Iso at o Valve-C os e HSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor.scram is initiated on a Hain Steam Isolation Valve-'Closure signal before the HSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient. However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux-High Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Code limits.That.is, the direct scram on position switches for ,MSIV closure events is not assumed in the overpressurization analysis.Additionally, MSIV closure is assumed in the transients analyzed in Reference 7 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow).The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the l-imits of 10 CFR 50.46.NSIV closure signals are initiated from position switches located on each of the eight HSIVs.Each HSIV has two position switches;one inputs to RPS trip system A while the other inputs to RPS trip system B.Thus, each RPS trip system receives an input from eight Hain Steam Isolation Valve-Closure channels, each consisting of one position switch.The logic for the Hain Steam Isolation Valve-Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.The Hain Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.(continued) BFN-UNIT 2 B 3.3-14 Amendment i Cl RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 5.ai Steam Isolat o'ive-Closu e (continued) Sixteen channels of the Hain Steam Isolation Valve-Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal.This Function is only required in NODE 1 since, with the HSIVs open and the heat generation rate high, a pressurization transient can occur if the NSIVs close.In NODE 2, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection. 6 e ess e-High pressure in the drywell could indicate a break in the RCPB.A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell.The Drywell Pressure-High Function is a secondary scram signal to Reactor Vessel Mater Level-Low, Level 3 for LOCA events inside the drywell.However, no credit is taken for a scram initiated from this Function for any of the DBAs analyzed in the FSAR.This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure.The Allowable Value was selected to be as low as possible and indicative of.a LOCA inside primary containment. Four channels of Drywell Pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required.to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.The Function is required in NODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents.(continued) BFN-UNIT 2 B 3.3-15 Amendment ik RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 7a b Scr'sc ar e Volume Water vel-The SDV receives the water displaced by the.motion of the CRD pistons during a reactor scram.Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram.The two types of Scram Discharge Volume Water Level-High Functions are an input to the RPS logic.No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the FSAR.However, they are retained to ensure the RPS remains OPERABLE.SDV water level is measured by two diverse methods.The level in each of the two SDVs is measured by two float type level switches and two thermal probes for a total of eight level signals.The outputs of these devices are arranged so that there is a signal from a level switch and a thermal probe to each RPS logic channel.The level measurement instrumentation satisfies the recommendations of Reference 8.The Allowable Value is chosen low enough to ensure that there, is sufficient volume in the SDV to accommodate the water from a full scram.Four channels of each type of Scram Discharge Volume Water Level-High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal.These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.8.Turbine Sto Valve-Closure Closure of the TSVs results in the loss of a-heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of (continued) BFN-UNIT 2 B 3.3-16 Amendment ik~Cl!5 RPS Instrumentation B'3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 8 rb e Sto V lve-C osure (continued) the transients that would result from the closure of these valves.The Turbine Stop Valve-Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 7.For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT)System, ensures that the MCPR SL is not exceeded.Turbine Stop Valve-Closure signals are initiated from position switches located on each of the four TSVs.Two independent position switches are associated with each stop valve.One of the two switches provides input to RPS trip system A;the other, to RPS trip system B.Thus, each RPS trip system receives an input from four Turbine Stop Valve-Closure channels, each consisting of one position switch.The logic for the Turbine Stop Valve-Closure Function is such that three or more TSVs must be closed to produce a scram.This Function must be-enabled at THERMAL POWER h 3K RTP.This is.normally accomplished automatically by pressure transmitters sensing turbine first stage pressure;therefore, opening the turbine bypass valves.may affect this function.The Turbine Stop Valve-Closure Allowable Value is selected to be high.enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient. Eight channels of Turbine Stop Valve-Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should close.This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is>3ÃRTP.This Function is not required when THERMAL'POWER is (3M'TP'ince the Reactor Vessel Steam Dome Pressure-High and the Average Power Range Monitor Fixed Neutron Flux-High Functions are.adequate to maintain the necessary safety margins.(continued) BFN-UNIT 2 B 3.3-17 Amendment il 0 RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 9.Turb e Control Valve ast Closure Tri 0 essure-ow Fast closure of the TCVs results in the loss of a heat sink that produces'reactor pressure, neutron flux, and heat flux transients that must be limited.Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves.The Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 7.For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the HCPR SL is not exceeded.Turbine Control Valve Fast Closure, Trip Oil Pressure-Low signals are initiated by the electrohydraulic control (EHC)fluid pressure at each control valve.One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel.This Function must be enabled at THERMAL POWER 2 3N'TP.This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure;therefore, opening the turbine bypass valves may affect this function.The Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Allowable Value is selected high enough to detect imminent TCV fast closure.Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument'failure will preclude a scram from this Function on a valid signal.This Function is required, consistent with the analysis assumptions, whenever THERHAL POWER is 2 30K RTP.This Function is not required when THERNL POWER is (30%RTP, since the Reactor Vessel Steam Dome Pressure-High and the Average Power Range Nonitor Fixed Neutron Flux-High Functions are adequate to maintain the necessary safety margins.(continued) BFN-UNIT 2 B 3.3-18 Amendment

RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 0 e ctor Mode Switch-S utdown os t The Reactor Mode Switch-Shutdown Position Function provides signals, via the manual scram logic channels, directly to the scram pilot.solenoid power circuits.These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically'redited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.Two channels of Reactor Mode Switch-Shutdown Position Function, with one channel in each trip system, are available and required to be OPERABLE.The Reactor Mode Switch-Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are'the MODES and other specified conditions when control rods are withdrawn. 11.ua Sera The Manual Scram push button channels provide signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits.These manual scram logic channels are redundant.to the automatic protective instrumentation channels and provide manual reactor tri'p capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.There is one Manual Scram push button channel for each of the two RPS manual scram logic channels.In order to cause a scram it is necessary that each channel in.both manual.scram trip systems be actuated.(continued) BFN-UNIT 2 B 3.3-19 Amendment ik~i 0 RPS Instrumentation 8 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY ll.Manual Scram (continued) There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.Two channels of Manual Scram with one channel in each manual scram trip system are available and required to be OPERABLE in NODES 1 and 2, and in NODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the NODES and other specified conditions when control rods are withdrawn. 12.RPS Channel Test Switches There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (Al, A2, Bl, and B2).These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip.When the RPS Channel Test Switch is placed in test, the associated scram logic channel is deenergized and OPERABILITY of the channel's scram contactors can be confirmed. The'RPS Channel Test Switches are not specifically credited in the accident analysis.However, because the Manual Scram Function at Browns Ferry Nuclear Plant is not configured the same as the generic model in Reference 9, the RPS Channel Test Switches are included in the analysis in Reference 11.Reference ll concludes that the Surveillance Frequency extensions for RPS functions, described in Reference. 9, are not affected by the difference in configuration since each automatic RPS channel has a test switch which is functionally the same as the manual scram switches in the generic model.Weekly testing of scram contactors is credited in Reference 9 with supporting the Surveillance Frequency extension of the RPS, functions. .There is no Al.lowable Value for this Function since the channels are mechanically actuated solely on the position of the switches.Four channels of the RPS Channel Test Switch Function with two channels in each trip system arranged in a one-out-of-two logic are available and required to be OPERABLE.The function is required in NODES 1 and 2, and in NODE 5 with (continued) BFN-UNIT 2 B 3.3-20 Amendment 0>>il RPS Instrumentation B 3.3.1.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABI L'I TY 12.PS C a ne Test Switches (continued) any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the NODES and other specified conditions when control rods are withdrawn. 'b 13.o Sc am lot A r Header ressur The Low Scram Pilot Air.Header Pressure trip performs the same function as the high water level in the scram discharge instrument volume for fast fill events in which the high level instrument response time may not be adequate.A fast fill event is postulated for certain degraded control air events in which the scram outlet valves unseat enough to allow 5 gpm per drive leakage into the scram discharge volume but not enough to cause rod insertion. The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram.Four channels of Low Scram Pilot Air Header Pressure Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.The Function is required in NODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and othe'pecified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.-However, the Required Actions for inoperable RPS instrumentation channels provide appropriate (continued) BFN-UNIT 2 B 3.3-21 Amendment 0 0 RPS Instrumentation B 3.3.1.1 ACTIONS (continued) compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (Ref.9)to permit restoration of any inoperable channel to OPERABLE status.However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.l, B.2, and C.l Bases).If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.l and A.2.Placing the inoperable channel in trip (or the associated trip system in trip)would conservatively.compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.Alternatively, if it is not desired to place the channel (or trip system)in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.~B.d 8.2 Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system.In this condition, provided at least one channel per trip system is OPERABLE;the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system.Required Actions B.l and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in Reference 9 for the 12 hour (continued) BFN-UNIT 2 B 3.3-22 Amendment il~Oi 0 RPS Instr umentation B 3.3.1.1 ACTIONS~d (3 dl Completion Time.Mithin the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in Reference 9, which justified a 12 hour allowable out of service time as presented in Condition A.The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what NODE the plant is in).If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.The 6 hour Completion Time is judged acceptable based on-the remaining capability to trip, the diversity of the sensors, available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.Alternately, if it is not desired to place the inoperable channels (or one trip system)in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken.Required Action C.l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (continued) BFN-UNIT 2 B 3.3-23 Amendment 0 0, RPS Instrumentation B 3.3.1.1 BASES ACTIONS~C.(continued)(or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.Required Action D.l directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and HODE or other specified condition dependent and may change as the ,Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.l.1 and G.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip)within the allowed Completion Time, the plant must be placed in a NODE or other specified condition in which the LCO does not apply.The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems.In addit'i'on, the Completion Time of Required Action E.l is consistent with the Completion Time provided in LCO 3.2.2,"HINIHUH CRITICAL POWER RATIO (HCPR);." If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in (continued) BFN-UNIT 2 B 3.3-24 Amendment I 0 il RPS Instrumentation B 3.3.1.1 BASES ACTIONS g,l (continued) trip)within the allowed Completion Time, the plant must be placed in a NODE or other specified condition in which the LCO does not apply.This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted.Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.SURVEILLANCE REQUIREMENTS As noted at the beginning of the SRs, the SRs for each RPS instrumentation Function are located in the SRs column of Table 3.3.1.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function, maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref.3)assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary. SR 3.3 1.1.Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK.is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or (continued) BFN-UNIT 2 B 3.3-25 Amendment I!Cl RPS Instrumentation B 3.3.1.1 SURVEILLANCE RE(UIREMENTS SR 3.3.1.1.1 (continued) something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key, to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency is based upon operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power.calculated from a heat balance.LCO 3.2.4,"Average Power Range Monitor (APRM)Gain and Setpoints," allows the APRMs to be reading greater, than actual THERMAL POWER to compensate for localized power peaking.When this adjustment is made, the requirement for the APRMs to indicate within 2%RTP of calculated power is modified.to require the APRMs to indicate within 2%RTP of calculated MFLPD.The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading, between performances of SR 3.'3.1.1.7.A restriction to satisfying this SR when<25%RTP is provided that requires the SR to be met only at a.25%RTP because it is diffi'cult to accurately maintain APRM indication of core THERMAL POWER consistent with, a heat balance when<25%RTP.At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR).At~25%RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2.A Note is provided, which allows an increase in THERMAL POWER above 25%if the 7 day Frequency is not met (continued) BFN-UNIT 2 B 3.3'-26 Amendment 0'~L' RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE RE(UIREMENTS SR 3.3.1.1.2 (continued) per SR 3.0.2.In this event, the SR must be performed within 12 hours after reaching or exceeding 25%RTP.Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.SR 3.3.1.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire.channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links.This allows entry.into MODE 2 if the 7 day Frequency is not met per SR 3.0.2.In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1.Twelve hours is based on operating experience and: in consideration of providing a reasonable time in which to complete the SR.A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref.9).SR 3.3.1.1.4 0 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel wi.l.l perform.the intended function.A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of Reference 9.(The RPS Channel Test Switch Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions'requencies.)(continued) BFN-UNIT 2 B 3.3-27 Amendment '~Ik RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) S 3.3 and S 3 3 1 6 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs.The overlap between IRHs and APRMs is of concern when reducing power into the IRM range.On power increases, the system design will prevent further increases (by initiating a rod block)if adequate overlap is not maintained. Overlap.between IRMs and APRHs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRH downscale rod block, or IRH upscale rod block.Overlap between SRHs and IRHs similarly exists when, prior to withdrawing the SRHs from.the fully inserted position, IRHs are above mid-seal'e on range 1 before SRHs have reached the upscale rod block.As noted, SR 3.3.1.1.6 is only required to be met during entry into NODE 2 from MODE 1.That is, after the overlap requirement has been met and indication has transitioned to the IRHs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).If overlap for a group of channels is not demonstrated (e.g., IRH/APRH overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current NODE or condition should be declared inoperable. A Frequency of 7 days is reasonable based on engineering judgment and the reliability.of the IRMs and APRHs.(continued) C BFN-UNIT 2 B 3.3-28 Amendment !l~i~0 RPS Instrumentation 8 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.1.1.7 LPRH gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)System.This establishes the relative local flux profile for appropriate representative input to the APRM System.The 1000 effective Full power hours Frequency is based on operating experience with LPRH sensitivity changes.SR 3.3.1.1.8 SR 3.3.1.1.12 and SR 3.3.1.1.16 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint.methodology. The 92 day Frequency of SR 3.3.1.1.8 is based on the rel.iability analysis of Reference 9.The 184 day Frequency of SR 3.3.1.1.16 for the scram pilot air header 1'ow pressure trip function is based on the functional reliability previously demonstrated by this function, the need for minimizing the radiation exposure associated with the functional testing of this function, and the increased risk to plant availability while the plant is in a half-scram condition during the performance of the functional testing versus the limited increase in reliability that would be obtained by the more frequent functional testing.The 18 month Frequency of SR 3.3.1.1.12 is based on the need to perform this Surveillance under the conditions that apply during a.plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.9 SR 3.3.1.1.10 and SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel (continued) BFN-UNIT 2 B 3.3-29 Amendment il~.i RPS Instrumentation B 3.3.1.1 SURVEILLANCE REgUIREHENTS SR 3.3.1.1.9 SR 3.3.1.1.10 and SR 3.3.1.1.13 (continued) adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. For the APRH Simulated Thermal Power-High Function, SR 3.3.1.1.9 also includes calibrating the associated recirculation loop flow channel.For HSIV-Closure, SDV Water Level-High (Float Switch), and TSV-Closure Functions, SR 3.3.1.1.13 also includes physical inspection and actuation of the switches.Note 1 to SR 3.3.1.1.9 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2)and the 1000 effective full power hours LPRH calibration against the TIPs (SR 3.3.1.1.7). A second Note for SR 3.3.1.1.9 is provided that requires the APRH and IRH SRs to be performed within 12 hours of entering HODE 2 from HODE 1.Testing of the NODE 2 APRH and IRH Functions cannot be performed in HODE 1 without utilizing jumpers, lifted leads, or movable 1;inks.This Note allows entry into NODE 2 from NODE 1 if the associated'requency is not met per SR 3.0.2.Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.'he Frequency of SR 3.3.1.1.9 is based upon the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.The Frequency of SR 3.3.1.1.10 is based upon the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.The Frequency of SR 3.3.1.1.13 is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of.equipment drift in the setpoint analysis.SR 3.3.1.1'.11 The Average Power Range Honitor Flow Biased Simulated Thermal Power-High Function uses the recirculation loop drive flows to vary the trip setpoint.This SR ensures that the total loop drive flow signals from the flow units used to vary the setpoint are appropriately compared to a (continued) BFN-UNIT 2 B 3.3-30 Amendment 0 RPS Instrumentation B 3.3.1.1 SURVEILLANCE RE(UIR EVENTS SR 3.3.1.1.11 (continued) The Frequency of,18 months is based on system design considerations which do not support flow unit bypass during operation. Thus, this calibration is performed during refueling outages.SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST.demonstrates the OPERABILITY of the required trip logic for a specific channel.The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.15 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions will not be inadvertently bypassed.when THERMAL POWER is)30%RTP.This involves calibration of the bypass channels.Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 30%RTP, either due to open main turbine bypass valve(s)or other reasons), then the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in.the nonbypass condition (Turbine Stop Valve-.Closure and Turbine Control Valve Fast Closure, Trip (continued) BFN-UNIT 2 B 3.3-31 Amendment il~ RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS. SR 3.3.1.1.'15 (continued) Oil Pressure-Low Functions are enabled),.this SR is met and the channel is considered OPERABLE.The Frequency of 18 months, is, based on engineering judgment and reliability of.the components. REFERENCES 1.FSAR, Section'..2. 2.FSAR, Chapter 14.3.NED0-23842,"Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.4..FSAR, Appendix N.5.FSAR, Section)4.6.2.6.FSAR, Section 6.5.7.FSAR, Section 14.5.8.P.Check (NRC)letter to,G..Lainas (NRC),,"BWR Scram Discharge System Safety Evaluation," December'1, 1980.9.NEDC-30851-P-A ,"Technical Specification Improvement Analyses for,BHR Reactor Protection System," March 1988..10.NRC No.93-102,"Final Policy Statement:on Technical Specification Improvements," July'23, 1993.11.NED-32-0286,"Technical Specification Improvement Analysis for Browns.Ferry Nuclear Plant, Unit 2," October 1995.BFN-UNIT'2 B 3.3-32 Amendment il SRH Instrumentation B 3.3.1.2 B 3.3 INSTRUMENTATION B 3.3.1.2 Source Range Monitor (SRH)Instrumentation BASES BACKGROUND The SRHs provide the operator with information relative to the neutron flux level at very low flux levels in the core.As such, the SRH indication is used by the operator to monitor the approach to criticality and determine when criticality is achieved.The SRHs are maintained fully inserted until the count rate is greater than a minimum allowed count rate (a control rod block is set at this condition). After SRH to intermediate range monitor (IRM)overlap is demonstrated (as required by SR 3.3.1.1.5), the SRMs are normally fully withdrawn from the core.The SRM subsystem of the Neutron Monitoring System (NHS), as described in Reference 1, consists of four channels.Each of the SRM channels can be bypassed, but only one at any given time, by the operation of a bypass switch.Each channel includes one detector that can be physically positioned in the core.Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning equipment, and electronics associated with the , various SRH functions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate.Each channel also includes indication, alarm, and control rod blocks.However, this LCO specifies OPERABILITY requirements only for the monitoring and indication functions of the SRHs.During refueling, shutdown, and low power operations, the primary indication of neutron flux levels is provided by the SRMs or special movable detectors connected to the normal SRH circuits.The SRHs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of subcritical multiplication that could be indicative of an approach to criticality. APPLICABLE SAFETY ANALYSES Prevention and mitigation of prompt reactivity excursions during refueling and low power operation is provided by LCO 3.9.l,"Refueling Equipment Interlocks"; LCO 3.1.1, (continued) BFN-UNIT 2 B 3.3-33 Amendment il ik SRM Instrumentation B 3.3.1.2 BASES APPLICABLE SAFETY ANALYSES (continued)"SHUTDOWN MARGIN (SDM)";LCO 3.3.1.1,"Reactor Protection System (RPS)Instrumentation"; IRM Neutron Flux-High and Average Power Range Monitor (APRM)Neutron Flux-High, Setdown Functions; and LCO 3.3.:2.1,"Control Rod Block Instrumentation." The SRMs have no safety function and, are not assumed to function during any FSAR design basis accident or transient analysis.However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical Specifications. LCO During startup in MODE 2, three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above.All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the core..In MODES 3 and 4, with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the core.In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core.Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueTed region containing at least one OPERABLE SRM is met.Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity (continued) BFN-UNIT 2 B 3.3-34 Amendment ~i il~ SRM Instrumentation B 3.3.1.2 LCO (continued) changes occurring in the reactor core.Because of the local nature of reactivity changes during refueling, adequate coverage is provided by requiring one SRM to be OPERABLE in the quadrant of the reactor core where CORE ALTERATIONS are being performed, and the other SRH to be OPERABLE in an adjacent quadrant containing fuel.These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS. Special movable detectors, according to footnote (c)of Table 3.3.1.2-1, may be used in place of the normal SRH nuclear detectors. These special detectors must be connected to the normal SRH circuits in the NMS, such that the applicable neutron flux indication can be generated. These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling. They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for SRMs.For an SRH channel to be considered OPERABLE, it must be providing neutron flux monitoring indication. APPLICABILITY The SRMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the IRMs being on scale on Range 3 to provide for neutron monitoring. In MODE 1, the APRHs provide adequate monitoring of reactivity changes in the core;therefore, the SRMs are not required.In MODE 2, with IRMs on Range 3 or above, the IRHs provide adequate monitoring and the SRHs are not required.ACTIONS A.l and B.l In MODE 2, with the IRMs on Range 2 or below, SRMs provide the means of monitoring core reactivity and criticality. With any number of the required SRHs inoperable, the ability to monitor neutron flux is degraded.Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.(continued) BFN-UNIT 2 B 3.3-35 Amendment i ik SRM Instrumentation B 3.3.1.2 ACTIONS A.l and B.1 (continued) Provided at least one SRH remains OPERABLE, Required Action A.1 allows 4 hours to restore the required SRHs to OPERABLE status.This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is sufficient time to take corrective actions to restore the required SRMs to OPERABLE status or to establish alternate IRM monitoring capability. During this time, control rod withdrawal and power increase is not precluded by this Required Action.Having the ability to monitor the core with at least one SRH, proceeding to IRH Range 3 or greater (with overlap required by SR 3.3.1.1.5), and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation. With three required SRMs inoperable, Required Action B.l allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes.Required Action A.l still applies and allows 4 hours to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately shut down, with no SRMs OPERABLE.C.1 In MODE 2, if the required number of SRHs is not restored to OPERABLE status within the allowed Completion Time, the reactor shall be placed in MODE 3.With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 in an orderly manner and without challenging plant systems.D.1 and D.2 With one or more required SRMs inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable (continued) BFN-UNIT 2 B 3.3-36 Amendment ili SRM Instrumentation B 3.3.1.2 BASES ACTIONS D.1 and 0.2 (continued) control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available. Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by maintaining a control rod block.The allowed Completion Time of 1 hour is sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the SRM occurring during this interval.E.l and E.2 With one or more required SRM inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded.CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the, reactor will be at its minimum reactivity given that fuel is present in the core.Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.'ction (once required to be initiated) to insert control rods must continue until all i'nsertable rods in core cells containing one or more fuel assemblies are inserted.SURVEILLANCE RE(UIREMENTS As noted at the beginning of the SRs, the SRs for each SRM Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1.SR 3.3.1.2.1 and SR 3.3.1.2.3 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar.parameter on another channel.It is based on the assumption that instrument channels (continued) BFN-UNIT 2 B 3.3-37 Amendment lli Q 0 SRH Instrumentation B 3.3.1.2 BASES SURVEILLANCE RE(U IREHENTS SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued) monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency of once every 12 hours for SR 3.3.1.2.1 is based on operating experience that demonstrates channel failure is rare.While in HODES 3 and 4, reactivity changes are not expected;therefore, the 12 hour Frequency is relaxed to 24 hours for SR 3.3.1.2.3. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core, when the fueled region encompasses more than one SRH, one SRH.is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRH must be in an adjacent quadrant containing fuel.Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is riot required to be met at other times in NODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRHs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE.In the event that only one SRH is required to be OPERABLE (when the fueled region encompasses only one SRH), per Table 3.3.1.2-1, footnote (b), only the a.portion.of this SR is required.Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRH.The 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities (continued) BFN-UNIT 2 B 3.3-38 Amendment ili il,' SRM Instrumentation B 3.3.1.2 SURVEILLANCE REQUIREMENTS SR 3.3.1.2.2 (continued) that include steps to ensure that the SRMs required by the LCO are in the proper quadrant.SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core.With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR.Therefore, allowances are made for loading sufficient"source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.To accomplish this, the SR is modified by Note 1 that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant.With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical.In addition, Note 2 states that this requirement does not have to be met during spiral unloading. If the core is being unloaded in this manner, the various core configurations encountered will not be critical.The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored whi.le core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours.SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly.SR 3.3.1.2.5 is required in MODE 5, and the 7 day Frequency ensures that the channels are OPERABLE whi.le core reactivity changes could be in progress.This Frequency is reasonable, based on (continued) BFN-UNIT 2 B 3.3-39 Amendment ik~il~ SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued) operating experience and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4.Since core reactivity changes do not normally take place in MODES 3 and 4 and core reactivity changes are due only to control rod movement in MODE 2, the Frequency has been extended from 7 days to 31 days.The 31 day Frequency is based on operating experience and on other Surveillances (such as CHANNEL CHECK)that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.Verification of the signal to noise ratio also ensures that the detectors are inserted to an acceptable operating level.In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector.Any count rate obtained while the detectors are fully withdrawn is assumed to be"noise" only.The Note to the Surveillance allows the Surveillance to be delayed, until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below).The SR must be performed within 12 hours after IRMs are on Range 2 or below.The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels.Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level.In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK)and the time required to perform the Surveillances. SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION at a Frequency of 92 days verifies the performance of the SRM detectors and (continued) BFN-UNIT 2 B 3.3-40 Amendment ik~il~ SRM Instrumentation B 3.3.1.2 SURVEILLANCE RE(UIREMENTS SR 3.3.1.2.7 (continued)'ssociated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status.The neutron detectors are excluded from the CHANNEL, CALIBRATION (Note 1)because they cannot readily be adjusted.The detectors are fission chambers that are des'igned to have a relatively constant sensiti.vity over the range and with an accuracy specified for a fixed useful life.Note 2 to the Surveillance allows the Surveillance to be delayed until entry into, the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below.The al.lowance to enter the Applicability with the 18 month Frequency.not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels.Although'the Surveillance could be performed while on IRM Range', the plant would not be expected to maintain steady state operation at this power level.In this event, the 12 hour'Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK)and the time required to.perform the Surveillances. REFERENCES 1.FSAR, Section 7.5.4.BF.N.-UNIT 2 B 3.3-41 Amendment ~i i ik Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes.Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM)provides protection for control rod withdrawal error events.'uring low power operations, control rod blocks from the rod worth minimizer (RWM)enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA).During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if local.ized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL)violation. The RBM supplies a trip signal to the Reactor Manual Control System (RHCS)to appropriately inhibit control rod withdrawal during power oper ation above the low power range setpoint.The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint.One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit.The RBM channel signal is generated by averaging a set of local power range monitor (LPRM)signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range monitor (APRM)channel assigned to each Reactor Protection System (RPS)trip system supplies a reference signal for the RBM channel in the same trip system.If the APRM is indicating less than the low power setpoint, the RBM is automatically bypassed.The RBM is also automatically bypassed if a peripheral control rod is selected (Ref.1).(continued) BFN-UNIT 2 B 3.3-42 Amendment il~0 Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND (continued) The purpose of the RWM is to control rod patterns during startup and'hutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to lOX RTP.The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA.Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence.The RWM determines the actual sequence'based position indication for each control rod.The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref..2).The RWM is a single channel system that provides input into both RMCS rod block circuits.With the reactor mode switch in.the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function, prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor.mode switch is required to be in the shutdown position.The reactor mode switch.has two channels, each inputting into a separate RHCS rod block circuit.A.rod block in either RMCS circuit will provide a control rod,block to all control, rods.APPLICABLE SAFETY.ANALYSES, LCO, and APPLICABILITY l.od Block Monitor The RBH is designed to prevent violation of the HCPR SL and the cladding lh plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE)event.The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3.A statisti'cal analysis of RWE events was performed to determine the RBH response for both channels for each event.From these responses, the fuel thermal'performance as a function of RBM Allowable Value was determined. Note that the RBH setpoint is flow-biased until implementation of ARTS improvements described in Reference 3.However, the generic RWE analysis in Reference 3 is currently applicable to establish required conditions for RBM OPERABILITY..(continued) BFN-UNIT 2 B 3.3-43 Amendment ik~ Control Rod Block Instrumentation B 3.3.2.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY d lock o tor (continued) The RBH Function satisfies Criterion 3 of the NRC Policy Statement (Ref.10).Two channels.of the RBH are required to be OPERABLE, with their setpoints within the appropriate Allowable Value to ensure that no single instrument failure can preclude a rod'lock from this Function.The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Oper ation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters.obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument error s (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration toler ances, instrument drift, and severe environmental effects (for channels that must function in, harsh environments as defined by 10 CFR 50.49)are accounted for.The RBH is assumed to mitigate the consequences of an RWE event when operating>2%'TP.Below this power level, the consequences of an RWE event will not exceed the HCPR SL and, therefore, the RBH is not required to be OPERABLE (Ref.3).When operating<9N RTP, analyses (Ref.3)have shown that with an initial HCPR h 1.70, no RWE event will result in exceeding the HCPR SL.Also, the analyses demonstrate that when operating at>9'TP with HCPR>1.40, no RWE event will result in exceeding the HCPR (continued) BFN-UNIT 2 B 3.3-44 Amendment Qi il~il Control Rod Block Instrumentation B 3.3.2.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 1 od 8 ock onitor (continued) SL (Ref.3).Therefore, under these conditions, the RBH is also not required to be OPERABLE.2.od Wo th inimize The RWM enforces the banked position withdrawal sequence (BPWS)to ensure that the initial conditions of the CRDA analysis are not violated.The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, and 7.The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6,"Rod Pattern Control." The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref.10).Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref.7).Special circumstances provided for in the Required Action of LCO 3.1.3,"Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods,,or to allow correction of a control rod pattern not in compliance with the BPWS.The RWM may be bypassed as.required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is (10%RTP.When THERMAL POWER is>I'll RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs.5 and 7).In MODES 3 and 4, all control rods are required to be inserted into the core;therefore, a CRDA cannot occur.In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDH ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.(continued) BFN-UNIT 2 B 3.3-45 Amendment ik~0 0 Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 3.e ctor Mode Sw t-Shutdow Posit'on During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed.The Reactor Mode Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref.10).Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required.There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE.During MODE 5 with the reactor.mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2,"Refuel Position One-Rod-Out Interlock")provides the required control rod withdrawal blocks.With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function;however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM.For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status.The Completion Time of 24 hours is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.(continued) BFN-UNIT 2 B 3.3-46 Amendment 0 Control Rod Block Instrumentation B 3.3;2.1 ACTIONS (continued) If Required Action A.l is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour.If both RBM channels are inoperable, the RBM is not capable of performing its intended function;thus, one channel must also be.placed in trip.This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.The 1 hour Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.C.l C..1 C...2 a d C.2.With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence.However, the overall reliability is reduced because a single operator error can'result in violating the control rod sequence.Therefore, control rod movement must be immediately suspended except by scram.Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM during withdrawal of one or more of the first 12 rods was not performed in the last 12 months.These requirements minimize the number of reactor startups initiated with the RWM inoperable. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs.and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2.Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator)or other qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer).(continued) BFN-UNIT'B 3.3-47 Amendment ~i 0 Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS C.2 C 1.d C.2.(continued) The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence.Required Action 0.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator)or other qualified member of the technical staff.The RWM may be bypassed under these conditions to allow the reactor shutdown to continue..1 nd With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function.However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable. In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1.Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted.Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.BFN-UNIT 2 B 3.3-48 (continued) Amendment ili i Control Rod Block Instrumentation B 3.3.2.1 SURVEILLANCE RE(UIREMENTS As noted at the beginning of the SRs, the SRs for each Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.The Surveillances are modified by a second Note (Note 2)to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the Surveill'ance, or expiration. of the 6 hour al,lowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref.9)assumption of the average time required to perform a channel.Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL-FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will per'form the intended function.It includes the Reactor Manual Control System input.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency. of 92 days is based on reliability analyses (Ref.8).SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function.The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs.This test is performed as soon as possible after the applicable conditions are entered.As noted in the SRs, SR 3.3.2.1.2 is not required to be.performed until 1 hour after any control rod is withdrawn at a 10%RTP in MODE 2.As noted, SR 3.3.2.1.3 is not required to be performed until (continued) BFN-'UNIT 2 B 3.3-49 Amendment il~0 0 Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued) 1 hour after THERMAL POWER is reduced to a 10%RTP in MODE 1.This allows entry into MODE 2 for SR 3.3.2.1.2, and THERMAL POWER reduction to a 10%RTP for SR 3.3.2.1.3, to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2.The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.The Frequencies are based on reliability analysis (Ref.8).SR 3.3.2.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted.to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.7. The Frequency is based upon the assumption of a 184 day calibration interval in, the determination of the magnitude of equipment drift in the setpoint analysis.SR 3.3.2.1.5 The RWM is automatically bypassed when power is above a specified value.The power level is determined from feedwater flow and steam flow signals.The automatic bypass setpoint must be verified periodically to be)10%RTP.If the RWM low power setpoint is nonconservative, then the RWN is considered inoperable. Alternately, the low power setpoint channel can be placed.in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the trip setpoint methodology utilized for the low power setpoint-channel.(continued) BFN-UNIT 2 B 3.3-50 Amendment il ll 0 Control Rod Block Instrumentation B 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) S 3.3.2.A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function.The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any control rod with'the reactor mode switch in the shutdown position.and verifying a control rod block occurs.As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.This allows entry into NODES 3 and 4 if the 18 month Frequency is not.met per SR 3.0.2.The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant.outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3 3.2.1 7.The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer.This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function.The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.REFERENCES 1.FSAR, Section 7.5.8.2.3. 2.FSAR, Section 7.16.5.3.1.k.(continued) BFN-UNIT 2 B 3.3-51 Amendment /Q~ik~il Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES (continued) 3.NEDC-32433P,"Haximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2 and 3," April 1995.4.NEDE-24011-P-A-US,"General Electrical Standard'pplication for Reload Fuel," Supplement for United States, (revision specified in the COLR).5."Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BMR Owners'Group, July 1986.6.NEDO-21231',"Banked Position Withdrawal Sequence," January 1977.7.NRC'ER,"Acceptance of Referencing, of Licensing Topical Report NEDE-24011-P-A,""General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27., 1987.8.NEDC-30851-P-A, Supplement 1,"Technical Specification Improvement Analysis for BWR Control Rod Block'nstrumentation," October 1988.'9.GENE-770-06-1,"Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.10.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-52 Amendment ~i 0 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 B 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.With excessive feedwater flow, the water level'n the reactor vessel ri'ses toward the high water level reference point, causing the trip of the three feedwater pump turbines and the main turbine.Reactor Vessel Water Level-High signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level in the reactor vessel (variable leg).Two channels of Reactor Vessel Water Level-High instrumentation per trip system are provided as input to a two-out-of-two initiation logic that trips the three feedwater pump turbines'and the main turbine.There are two trip systems, either of which will initiate a trip.The channels include electronic equipment (e.g., trip units)that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a main feedwater and turbine trip signal to the trip logic.A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel.A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.APPLICABLE SAFETY ANALYSES The feedwater and main turbine high water level trip instrumentation is assumed to be capable of, providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref.I).The reactor vessel high water level trip indirectly initiates a reactor, scram from the main turbine trip (above 30%RTP)and trips the feedwater pumps, thereby terminating (continued) BFN-UNIT 2 B 3.3-53 Amendment il~i ik Feedwater and Hain Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued) the event.The reactor scram mitigates the reduction in HCPR.Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.3).LCO The LCO requires two channels of the Reactor Vessel Water Level-High instrumentation per trip system to be OPERABLE to ensure that no single instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Reactor Vessel Water Level-High, signal.Both channels in either trip system are needed to provide trip signals in order for the feedwater and main turbine trips to occur.Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event.The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. 'ominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive, CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, (continued) BFN-UNIT 2 B 3.3-54 Amendment O~ll Feedwater and Hain Turbine High Mater Level Trip Instrumentation B 3.3.2.2 BASES LCO (continued) instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to'be OPERABLE at a 25%RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1%plastic strain, limit are not violated during the feedwater controller failure, maximum demand event.As discussed in the Bases for LCO 3.2.1,"Average Planar Linear Heat Generation Rate (APLHGR)," and LCO 3.2.2,"MINIMUM CRITICAL POWER RATIO (HCPR)," sufficient margin to these limi.ts exists below 25%'RTP;therefore, these requirements are only necessary when operating at or above this power level.A Note.has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition'has.been entered, subsequent divisions, 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. Section 1.3 also specifies'that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip instrumentation channel.A.1'ith one channel inoperable in one trip system, the*remaining two OPERABLE channels in the other trip system can provide the required trip signal.However, overall instrumentation reliability is reduced because a single failure in one of the two channels of that trip system (continued) BFN-UNIT 2 B 3.3-55 Amendment il~ll'l Feedwater and Hain Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS A.l (continued) concurrent with feedwater controller failure, maximum demand event, may result in the instrumentation not being able to perform its intended function.Therefore, continued operation is only allowed for a limited time with one channel inoperable. If the inoperable channel cannot be restored to OPERABLE status within the Completion Time, the channel must be placed in the tripped condition per Required Action A.l.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would.result in a feedwater or main turbine trip), Condition C must be entered and its Required Action taken.The Completion Time of 7 days is based on the low probability of the event occurring coincident with a single failure in a'remaining OPERABLE channel.B.l With one or more channels inoperable in each trip system, the feedwater and main turbine high water level trip instrumentation cannot perform its design function (feedwater and main turbine high water level trip capability is not maintained). Therefore, continued operation is only permitted for a 2 hour period, during which feedwater and main turbine high water level trip capability must be restored.The.trip capability is considered maintained when sufficient channels are OPERABLE or in trip.such that the feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal.This requires that two channels in one trip system be OPERABLE or in trip.If the required channels cannot be restored to OPERABLE status or placed in trip, Condition C must be entered and its Required Action taken.The 2 hour Completion Time is sufficient for the operator to take corrective action, and takes into account the 1-ikelihood of an event requiring actuation of feedwater and main turbine high water level trip instrumentation occurring (continued) BFN-UNIT 2 8 3.3-56 Amendment ~i i Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS B.1 (continued) during this period.It is also consistent with the 2 hour Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation. C.1 With the required channels not restored to OPERABLE status or placed in trip, THERMAL POWER must be reduced to<25%RTP within 4 hours.As discussed in the Applicability section of the Bases, operation below 25%RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event.The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to<25%RTP from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE RE(UIREMENTS The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref.2)assumption of the average time required to perform channel Survei.llance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the feedwater pump turbines and main turbine will trip when necessary. SR 3.3.2.2.1 Performance-of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A (continued) BFN-UNIT 2 B 3.3-57 Amendment ~i Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.2.1 (continued) CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits.The Frequency is based on operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.2.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analysis (Ref.2).SR 3.3.2.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive (continued) BFN-UNIT 2 B 3.3-58 Amendment il~0 Feedwater and Hain Turbine High Mater Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.2.3 (continued) calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.SR 3.3.2.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.The system functional test of the feedwater and main turbine valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function.Therefore, if a valve is incapable of operating, the associated instrumentation would also be inoperable. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. REFERENCES 1.FSAR, Section 14.5.7.2.GENE-770-06-1,"Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications," February 1991.3.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-59 Amendment 0 PAM Instrumentation B 3.3.3.1 B 3.3.3.1 Post Accident Monitoring (PAM)Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events.The instruments that monitor these variables are designated as Type A, Category 1, and non-Type A, Category 1, in accordance with Regulatory Guide 1.97 (Ref.1).The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident.This capability is consistent with the recommendations of Reference 1.APPLICABLE SAFETY ANALYSES The PAM'nstrumentation LCO ensures the OPERABILITY of Regulatory Guide 1.97, Type A variables so that the control room operating staff can:~Perform the diagnosis specified in the Emergency Operating Instructions (EOIs).These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and~Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.The PAM instrumentation LCO also ensures OPERABILITY of Category 1, non-Type A', variables so that the control room operating staff can:~Determine whether systems important to safety are performing their intended functions; j (continued) BFN-UNIT 2 8 3.3-60 Amendment lli 0 PAN Instrumentation B 3.3.3.1 BASES APPLICABLE SAFETY ANALYSES (continued) ~Determine the potential for causing a gross breach of the barriers to radioactivity release;~Determine whether a gross breach of a barrier has occurred;and~Initiate action necessary to protect the public and for an estimate of the magnitude of, any impending threat.The plant specific Regulatory Guide 1.97 Analysis (Ref.2).documents the process that identified Type A and Category 1, non-Type A,.variables. Accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of the NRC Policy Statement (Ref.6).Category 1, non-Type A, instrumentation is retained in Technical Specifications (TS)because they are intended to assist operators in minimizing the consequences of accidents. Therefore, these Category 1 variables are important for reducing public risk.LCO LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the-plant to, and maintain it in, a safe condition following that accident.Furthermore, provision of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. The exception to the two channel requirement is primary containment isolation valve (PCIV)position.In this case, the important information is the status of the primary containment penetrations. The LCO requires one position indicator for each active (e.g., automatic) PCIV.This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive valve or via system boundary.status.If a normally active PCIV is known to be closed and deactivated, position indication is not needed to determine status.Therefore, the position (continued) BFN-UNIT 2 B 3.3-61 Amendment ~i il~ PAM Instrumentation B 3.3.3.1 BASES LCO (continued) indication for closed and deactivated.valves is not required to be OPERABLE.The following list is a discussion of the specified instrument Functions listed in Table 3.3.3.1-1. 1.Reactor Steam Dome Pressure Reactor steam dome pressure is a Category)variable provided to support monitoring of Reactor Coolant System (RCS)integrity and to verify operation of the Emergency Core Cooling Systems (ECCS).Two independent pressure transmitters with a range of 0 psig to 1200 psig monitor pressure.Wide range indicators are the primary indication used by the operator during an accident.Therefore, the PAM Specification deals specifically with this portion of the instrument channel.2.Reactor Vessel Water Level Reactor vessel water level is a Category I variable provided to support monitoring of core cooling and to verify operation of the ECCS.Two different range water level channels (Emergency Systems and Post-accident Flood Range)provide the PAM Reactor Vessel Water Level Functions. The water level channels measure from I/3 of the core height to 221 inches above the top of the active fuel.Water 1'evel is measured by two independent differential pressure transmitters for each required channel.The output from these channels is indicated on two independent indicators, which is the primary indication used by the operator during an accident.Therefore, the PAM Specification deals specifically with this portion of the instrument channel.The reactor vessel water level instruments are not compensated for variation in reactor water density.Function 2.a is calibrated to be most accurate at operational pressure and temperature while Function 2.b is calibrated to be most accurate for accident conditions. {continued) BFN-UNIT 2 B 3.3-62 Amendment ik~il>>O~ PAM.Instrumentation B 3.3.3.1 LCO (continued) 3.Su ression Pool Mater Level Suppression pool water level is a Category 1 variable pr'ovided to detect a breach in the reactor coolant pressure boundary (RCPB).This variable is also used to verify and provide long term surveillance of ECCS function.The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS.The wide range water level indicators monitor the suppression pool water level from two feet from the bottom of the pool to five feet above normal water level.Two wide range suppression pool water level signals are transmitted from separate differential pressure transmitters and are continuously recorded and displayed on one recorder and one indicator in the control room.The recorder and indicator are the primary indication used by the operator during an accident.Therefore, the PAM Specification deals specifically with this portion of the instrument channel.4.Or well Pressure Drywell pressure is a Category 1 variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two different ranges of drywell pressure channels (normal and wide range)receive signals that are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders and two control room indicators. These recorders and indicators are the primary indication used'by the operator during an accident.Therefore, the PAM Specification deals specifically with this portion of the instrument channel.5.Primar Containment Area Radiation Hi h Ran e Primary containment area radiation (high range)is provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans.Two high range primary containment area radiation signals are transmitted from separate radiation detectors and are continuously recorded and displayed on two control room recorders. These recorders are the primary indication used (continued) BFN-UNIT 2 B 3.3-63 Amendment il~il~ PAM Instrumentation B 3.3.3.1 BASES LCO 5.Primar Containment Area Radiation Hi h Ran e (continued) by the operator during an accident.Therefore, the PAM Specification deals specifically with this portion of the instrument channel.6.Primar Containment Isolation Valve PCIV Position PCIV position is provided for verification of containment integrity. In the case of PC IV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two total channels of PCIV position indication for a penetration flow path with two active valves.For containment penetrations with only one active PCIV having control room indication, Note (b)requires a single channel of valve position indication to be OPERABLE.This is sufficient.to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status.If a penetration flow path is isolated, position indication for the PCIV(s)in the associated penetration flow path is not needed to determine status.Therefore, the position indication for valves in an isolated penetration flow path is not required to be OPERABLE.The indication for each PCIV consists of green and red indicator lights that illuminate to indicate whether the PCIV is fully open, fully closed, or in a mid-position. Therefore, the PAM specification deals specifically with this portion of the instrument channel.7.Dr well and Torus H dro en Anal zers Drywell and torus hydrogen analyzers are Category 1 instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach.The drywell and torus hydrogen concentration recorders al]ow the operators to detect trends in hydrogen concentration in sufficient time to initiate (continued) BFN-UNIT 2 B 3.3-64 Amendment ~i ik~ PAM Instrumentation B 3.3.3.1 BASES LCO 7.Dr well and Torus H dro en Anal zers (continued) containment atmospheric dilution if containment atmosphere approaches combustible limits.Hydrogen concentration indication is also important in verifying the adequacy of mitigating actions.High hydrogen concentration is measured by two independent analyzers and continuously recorded and displayed on one control room recorder and one control room.indicator. The analyzers have the capability for sampling both the drywell and the torus.These indicators are the primary indication used by the operator during an accident.Therefore, the PAN Specification deals specifically with this portion of the instrument channel.8.Su ression Pool Water Tem erature Suppression pool water temperature is a Category 1 variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach.The suppression pool water temperature instrumentation allows operators to detect trends in suppression pool water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool.Sixteen temperature sensors are arranged in two groups of two independent and redundant. channels, located such that they are sufficient to provide a reasonable measure of bulk pool temperature. The outputs for the sensors are recorded on two independent recorders in the control room.These recorders are the primary indication used by the operator during an accident.Therefore, the PAN Specification deals specifically with this portion of the instrument channels.9.Dr well Atmos here Tem erature Drywell atmosphere temperature is a Category 1 vari able provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach.Two wide range drywell atmosphere temperature signals are transmitted from separate temperature transmitters and are continuously recorded and displayed on one control room recorder and one control'oom indicator. The recorder and indicator are the primary indications used by the operator during an accident.(continued) BFN-UNIT 2 B 3.3-65 Amendment 4I~ik~0' PAM Instrumentation B 3.3.3.1 BASES LCO 9.Dr well Atmos here Tem erature (continued) Therefore, the PAM Specification deals specifically with this portion of the instrument channel.APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1 and 2.These variables are related to the diagnosis and preplanned actions required to mitigate DBAs.The applicable DBAs are, assumed to occur in MODES 1 and 2.In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low;therefore, PAM instrumentation is not required to be OPERABLE in these MODES.ACTIONS Note 1 has been added to the ACTIONS to exclude the MODE change restriction of LCO 3.0.4.This exception allows entry into the applicable MODE while, relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown.This exception is acceptable due to the passive function of the instruments, the operator's ability to diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments. Notes 2 and 3 have been provided to modify the ACTIONS related to PAM instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, Note 2 has been provided to allow separate Condition entry for each inoperable PAM Function.Note 3 has been provided for Function 6 to allow separate Condition entry for each penetration flow path.(continued) BFN-UNIT 2 B 3.3-66 Amendment Cli ggi ll~ PAM Instrumentation B 3.3.3.1 BASES ACTIONS (continued) A.1 When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status.within 30 days.The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAN instrumentation during this interval.B.l If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with.Specification 5.6.6, which requires a written report to be submitted to the NRC.This report discusses the alternate method of monitoring, the results of the root cause evaluation of the inoperability, and i'dentifies proposed restorative actions.This action is appropriate in lieu of a shutdown requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this-instrumentation. C.1 When one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days.The Completion Time of 7 days.is based on the relatively low probability of an event requiring PAN instrument operation and the availabil-ity of alternate means to obtain the required information. Continuous operation with two required channels in'operable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM'nstrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the (continued) BFN-UNIT 2 B 3.3-67 Amendment ili ili ili PAN Instrumentation B 3.3.3.1 BASES ACTIONS C.1 (continued) PAM Function will be in a degraded condition should an accident occur.Condition C is modified by a Note that excludes hydrogen monitor channels.Condition D provides appropriate Required Actions for two inoperable hydrogen monitor channels.D.1 When two hydrogen monitor channels are inoperable, one hydrogen monitor channel must be restored to OPERABLE status within 72 hours.The 72 hour Completion Time is based on the low probability of the occurrence of a LOCA, that would generate hydrogen in amounts capable of exceeding the flammability limit;and the length of, time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit.E.l This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met any Required Action of Condi,tion C or 0, as applicable, and the associated Completion Time has expired, Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition. F.1 For the, majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C or D are not met, the plant must be brought to a NODE in which the LCO not apply.To achieve this status, the plant must be brought to at least MODE 3 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.(continued) BFN-UNIT 2 B 3.3-68 Amendment i ili il~ PAN Instrumentation B 3.3.3.1 BASES ACTIONS (continued) G.l Since alternate means of monitoring primary containment area radiation have been developed and tested, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6.These alternate means may be temporarily installed if the normal PAN channel cannot be restored to OPERABLE status within the allotted time.The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAN channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAN channels.SURVEILLANCE REQUIREMENTS SR 3.3.3.1.1 Performance of the CHANNEL CHECK for each required PAN instrumentation channel once every 31 days ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter. indicated on one channel against a similar parameter on other.channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between instrument channels could be an indication of excessive"instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrument channels should be compared to each other or to other containment radiation monitoring instrumentation. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.The Frequency of 31 days is based upon plant operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of (continued) BFN-UNIT 2 B 3.3-69 Amendment O~il PAN Instrumentation B 3.3.3.1 BASES SURVEILLANCE RE(UIRENENTS SR 3.3.3.1.1 (continued) a given Function in any 31 day interval is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the required channels of this LCO.SR 3.3.3.1.2 and SR 3.3.3.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor.The test verifies the channel responds to measured parameter with the necessary range and accuracy.For the PCIV position function, the CHANNEL CALIBRATION consists of verifying the remote indications conform to actual valve positions. The 92 day Frequency for CHANNEL CALIBRATION of the Drywell and Torus Hydrogen Analyzer is based on operating experience and vendor recommendations. The 18 month Frequency for CHANNEL CALIBRATION of all other PAN instrumentation in Table 3.3.3.1-1 is based on operating experience and consistency with BFN refueling cycles.REFERENCES 1.Regulatory Guide 1.97,"Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, Hay 1983.2.TVA Letter from L.M.Hills to H.R.Denton (NRC)dated April 30, 1984.3.NRC Letter from S.C.Black to S.A.White (TVA), NRC Regulatory Guide 1.97 SER letter, dated June 23, 1988.4.TVA General Design Criteria No.BFN-50-7307, Revision 4,"Post-Accident Honitoring," dated June 22, 1993.5.NRC Letter from Joseph F.Williams to Oliver D.Kingsley, Jr.,"Regulatory Guide 1.97-Boiling Water Reactor Neutron Flux Monitoring For the Browns Ferry Nuclear Plant, Units 1, 2, and 3," dated May 3, 1994.6.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-70 Amendment 0 Backup Control System B 3.3.3.2 B 3.3 INSTRUMENTATION B 3.3.3.2 Backup Control System BASES BACKGROUND The Backup Control System provides the control room operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the control room.This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3.With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC)System, the safety/relief valves, and the Residual Heat Removal System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC and the ability to operate the RHR System for decay heat removal from outside the control room allow extended operation in MODE 3.In the event that the control room becomes inaccessible, the operators can establish control at the backup control panel and place and maintain the plant in MODE 3..Not all controls and necessary transfer switches are located at the backup control panel.Some controls and transfer switches will have to be operated locally at the switchgear, motor control panels, or other local stations.The plant automatically reaches MODE 3 following a plant shutdown and can be maintained safely in MODE 3 for an extended period of time.The OPERABILITY of the Backup Control System control and instrumentation Functions ensures that there is sufficient information available on selected plant parameters to place and maintain the plant in MODE 3 should the control room become inaccessible. APPLICABLE SAFETY ANALYSES The Backup Control System is required to provide equipment at appropriate lo'cations outside the control room with a design capability to promptly shut down the reactor to MODE 3, including the necessary instrumentation and controls, to maintain the plant in a safe condition in MODE 3.(continued) BFN-UNIT 2 B 3.3-71 Amendment 4I Backup Control System B 3.3.3.2 BASES APPLICABLE SAFETY ANALYSES (continued) The criteria governing the design and the specific system requirements of the Backup Control System are located in 10 CFR 50, Appendix A, GDC 19 (Ref.1)and Reference 2.The Backup Control System.,is considered an important contributor to reducing the risk of accidents; as such, it meets Criterion 4 of the NRC Policy Statement (Ref.3).LCO The Backup Control System LCO provides the requirements for the OPERABILITY of the instrumentation and controls necessary to place and maintain the plant in NODE 3 from a location other than the control room.The instrumentation and controls typically required are listed in Table B 3.3.3.2-1. The controls, instrumentation, and transfer switches are those required for:~Reactor pressure vessel (RPV)pressure control;~Decay heat removal;~RPV inventory control;and~Safety support systems, for the above functions, including Residual Heat Removal (RHR)Service Water, Emergency Equipment Cooling Water, and onsite power, including the diesel generators. The Backup Control System is OPERABLE if all instrument and control channels needed to support the backup control function are OPERABLE.In some cases, Table B 3.3.3.2-1 may indicate that the required information or control capability is available from several'lternate sources.In these cases, the Backup Control System is OPERABLE as long as one channel of any of the alternate information or control sources for each Function is OPERABLE.The Backup Control System instruments and control circuits covered by this LCO do not need to be energized to be considered OPERABLE.This LCO is intended to ensure that the instruments and control circuits will be OPERABLE if plant conditions require that the Backup Control System be placed in operation. I BFN-UNIT 2 8 3.3-72 (continued) Amendment /gi Backup Control System B 3.3.3.2 BASES (continued) APPLICABILITY The Backup Control System LCO is applicable in MODES 1'nd 2.This is required so that the plant can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room.This LCO is not applicable in MODES 3, 4, and 5.In these MODES, the plant is already subcritical and in a condition of reduced Reactor Coolant System energy.Under these conditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control becomes unavailable. Consequently, the TS do not require OPERABILITY in MODES 3, 4, and 5.ACTIONS A Note is included that excludes the MODE change restriction of LCO 3.0.4.This exception al.lows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a plant shutdown.This exception is acceptable due to the low probability of an event requiring this system.Note 2 has been provided to modify the ACTIONS related to Backup Control System Functions. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable Backup Control System Functions provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable Backup Control System Function.A.1 Condition A addresses the situation where one or more required Functions of the Backup Control System is inoperable. This includes any Function listed in Table B.3.3.3.2-1, as well as the control and transfer switches.(continued) BFN-UNIT 2 B 3.3-73 Amendment ili il~il Backup Control System B 3.3.3.2 ACTIONS A.1 (continued) The Required Action is to restore the Function to OPERABLE status within 30 days.The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.If the Required Action and associated Completion Time of Condition A are not met, 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 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE RE(UI REMENTS SR 3.3.3.2.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL'ALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized.(continued) BFN-UNIT 2 B 3.3-74 Amendment ili if'l Backup Control System 8 3.3.3.2 SURVEILLANCE REQUIREMENTS SR 3.3.3.2.1 (continued) The Frequency is based upon plant operating experience that demonstrates channel failure is rare.SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Backup Control System transfer switch and control circuit performs the intended function.This verification is performed from the backup control panel and locally, as appropriate. Operation of the equipment from the backup control panel is not necessary. The Surveillance can be satisfied by performance of a continuity check.This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the backup control panel and the local control stations.Operating experience demonstrates that Backup Control System control channels usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.3.2.3 and SR 3.3.3.2.4 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.The test verifies the channel responds to measured parameter values with the necessary range and accuracy.The Frequency of SR 3.3.3.2.3 is based upon the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.The 18 month Frequency of SR 3.3.3.2.4 is based upon operating experience and consistency with the typical industry refueling cycle.REFERENCES l.10 GFR 50, Appendix A, GDC 19.2.FSAR Section 7.18.3.NRC No..93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-75 Amendment il~O~ Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 1 of 4)Backu Control System Instrumentation and Controls FUNCTION Instrument Parameter REQUIRED NUMBER OF CHANNELS 1.Reactor Water Level Indication 2.Reactor Pressure Indication 3.4~5.6.7.Suppression Pool Temperature Indication Suppression Pool Level Indication Drywell Pressure Indication Drywell Temperature Indication EECW Flow Indication 8.RCIC Flow Indication 9.10.RCIC Turbine Speed Indication RCIC Turbine Trip Alarm RCIC Turbine Bearing Oil High Temperature Alarm.1 1 1 1 1 1 2 (1/Header) 1 1 1 1 16.17.RHRSW Discharge Valves for RHR Loop I Heat Exchangers RCW Pumps 1D and 3D (Trip Function Only)Transfer Control Parameter 12.MSRV Transfer&Control 13.MSIV Transfer&Control (Closure Only)14.Main Steam Drain Line Isolation Valves 15.RHRSW Pumps 3 (1/MSRV)8 (1/MSIV)2 (1/valve)12 (1/pump)2 (1/valve)2 (1/pump)BFN-UNIT 2 B 3.3-76 Amendment Oi Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 2 of 4)Backup Control System Instrumentation and Controls FUNCTION Transfer Control Parameter continued REQUIRED NUHBER OF CHANNELS 23.24.Recirculation Pump Discharge Valve (RHR Loop I LPCI)RWCU Drain to Hain Condenser Hotwell Isolation Valve 25.RWCU Drain to Radwaste Isolation Valve 26.RBCCW Pump Controls 27.Drywell Cooler RBCCW Flow Control Valves 28.Drywell Cooler Fan Controls 29.RHR Shutdown Cooling Inboard Containment Isolation Valve 30.RHR Shutdown Cooling Outboard Containment Isolation Valve 31.RCIC Steam Supply Isolation Valves 32.RCIC Steam Pot Drain Line Steam Trap Bypass 18.4-kV Fire Pumps A, B, and C 19.Recirculation System Sample Line Isolation Valves 20.EECW Sectionalizing Valves 21.RHRSW to EECW Motor-Operated Crosstie Valves 22.EECW Supply to RBCCW Heat Exchangers 3 (1/pump)2 (1/valve)8 (1/val ve)2 (1/valve)6 (1/val ve)1 1 2 (1/pump)10 (1/cool er)10 (1/cool er)1 2 (1/valve)1 BFN-UNIT 2 B 3.3-77 Amendment ili ll 4 ggi Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 3 of 4)Backup Control System Instrumentation and Controls FUNCTION Transfer Control Parameter continued REQUIRED NUNBER OF CHANNELS 33.RCIC Steam Pot Drain to Main Condenser Isolation 34.RCIC Drain to Radwaste Isolation 38.39.40.41.42.43.44.45.46.47.RCIC Pump Suction From Condensate Storage Tank RCIC Lube Oil Cooler Cooling Water Supply RCIC Pump Minimum Flow Bypass RCIC Pump Discharge RCIC Test Return to'Condensate Storage Tank RCIC Injection Valve to Reactor Vessel RCIC Barometric Condenser Condensate Pump RCIC Barometric Condenser Vacuum Pump HPCI Turbine Steam Supply Valve (Isolation Function Only)RHR Pump Controls 48.RHR Loop I Motor Operated Val'ves 35.RCIC Turbine Steam Supply Valve 36.RCIC Turbine Stop Valve 37.RCIC Pump Suction From Suppression Pool 1 (1 switch for 2 valves)1 (1 switch for 2 valves)1 1 2 (1/val ve)1 4 (1/pump)17 (1/val ve)BFN-UNIT 2 B 3.3-78 Amendment ~i i Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 4 of 4)Backup Control System Instrumentation and Controls FUNCTION Transfer Control Parameter continued REQUIRED NUHBER OF CHANNELS 49.50.51.-Core Spray Pumps (Trip 8 Lock-out Function Only)CRD Pump 1B CRD Pump Discharge Valves 52.Scram Discharge Volume Isolation Pilot Valve (1/pump)1 2 (1/valve)1 BFN-UNIT 2 8 3.3-79 Amendment il~3g~il EOC-RPT Instrumentation B 3.3.4.1 B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)Instrumentation BASES BACKGROUND The EOC-RPT instrumentation initiates a recirculation pump trip (RPT)to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal HCPR Safety Limits (SLs).The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV)Fast Closure, Trip Oil Pressure-Low or Turbine Stop Valve (TSV)-Closure.The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity. The EOC-RPT instrumentation, as shown in Reference 1, is composed of sensors that detect initiation of closure of the TSVs or fast closure of the TCVs, combined with relays, logic circuits, and fast acting circuit breakers that interrupt power from the recirculation pump motor generator (MG)set generators to each of the recirculation pump motors.The channels include electronic equipment (e.g., trip relays)that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an EOC-RPT signal to the trip logic.When the RPT breakers trip open, the recirculation pumps coast down under their own inertia.The EOC-RPT has two identical trip systems, either of which can actuate an RPT.Each EOC-RPT trip system is a two-out-of-two logic for each Function;thus, either,two TSV-Closure or two TCV Fast Closure, Trip Oil Pressure-Low signals are required for a trip system to actuate.If either trip system actuates, both recirculation pumps will trip.There are two EOC-RPT breakers in series per recirculation pump.One trip system trips one of the two EOC-RPT breakers for each recirculation (continued) BFN-UNIT 2 8 3.3-80 Amendment il~iS~ik EOC-RPT Instrumentation B 3.3.

4.1 BACKGROUND

(continued) pump, and the second trip system trips the other EOC-RPT breaker for each recirculation pump.APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The TSV-Closure and the TCV Fast Closure, Trip Oil Pressure-Low Functions are designed to trip the recirculation pumps in the event of a turbine trip or generator load rejection to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the NCPR SL, The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2, 3, and 4.To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs.The combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting in an increased margin to the NCPR SL.Alternatively,, MCPR limits for an inoperable EOC-RPT, as specified in the COLR, are sufficient to prevent violation of the NCPR Safety Limit.The EOC-RPT function is automatically disabled when turbine first stage pressure is<30%RTP.EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.6).The OPERABILITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1'.3.The setpoint is calibrated consistent with applicable setpoint methodology assumpti'ons (nominal trip setpoint). Channel OPERABILITY also includes the associated EOC-RPT breakers.Each channel (including the associated EOC-RPT breakers)must also respond within its assumed response time.Allowable Values are specified for each EOC-RPT Function specified in the LCO.Nominal trip setpoints are specified in the setpoint calculations. A channel is inoperable if its actual trip setpoint is not within its required Al-lowable Value.The nominal setpoints are selected to ensure that the setpoints-do not exceed the Allowable Value (continued) BFN-UNIT 2 B 3,3-81 Amendment ~i II EOC-RPT Instrumentation B 3.3.4.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit assumed in the transient and accident analysis in order to account for instrument uncertainties appropriate to the Function.Trip setpoints are those predetermined values of output at which an action should take place.The'setpoints are compared to the actual process parameter (e.g., TSV position), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.Alternatively, since this instrumentation protects against a MCPR SL violation, with the instrumentation inoperable, modifications to the MCPR limits (LCO 3.2.2)may be applied to allow this LCO to be met.The MCPR penalty for the EOC-RPT inoperable condition is specified in the COLR.Turbine Sto Valve-Closure Closure of the TSVs and a main turbine trip result in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.Therefore, an RPT is initiated on TSV-Closure in anticipation of the transients that would result from closure of these valves.EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.(continued) BFN-UNIT 2 B 3.3-82 Amendment il~ EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Turbine Sto Valve-Closure (continued) Closure of the TSVs is determined by measuring the position of each valve.There are two separate position switches associated with each stop valve, the signal from each switch being assigned to a separate trip channel.The logic for the TSV-Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT.This Function must be enabled at THERMAL POWER a 30%RTP.This is normally.accomplished automatically by pressure transmitters sensing turbine first stage pressure;therefore, opening the turbine bypass valves may affect this function.Four channels of TSV-Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT, from this Function on a valid signal.The TSV-Closure Allowable Value is selected to detect imminent TSV closure.This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is w 30%RTP.Below 30%RTP, the, Reactor Vessel Steam Dome Pressure-High and the Average Power Range Monitor (APRM)Fixed Neutron Flux-High Functions"of the Reactor Protection System (RPS)are adequate to maintain the necessary margin to the MCPR Safety.Limit.Turbine Control Valve Fast Closure Tri Oil Pressure-Low Fast closure of the TCVs during a generator load rejection results in the loss of a heat.sink that produces reactor pressure, neutron flux, and heat flux transients. that must be limited.Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure-Low in anticipation of the transients that would result from the closure of these valves.The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. Fast closure of the TCVs is determined by measuring the electrohydraulic control.fluid pressure at each control valve.There is one pressure transmitter associated with each control valve, and the signal from each transmitter is assigned to a separate trip channel.The logic for the TCV Fast Closure, Trip Oil Pressure-Low Function is such that two or more TCVs must be closed (pressure transmitter trips)(continued) BFN-UNIT 2 B 3.3-83 Amendment il~gg~ EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Turbine Control Valve Fast Closure Tri Oil Pressure-Low (continued) to produce an EOC-RPT.This Function must be enabled at THERMAL POWER a 30%RTP.This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure;therefore, opening the turbine bypass valves may affect this function.Four channels of TCV Fast Closure, Trip Oil Pressure-Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal.The TCV Fast Closure, Trip Oil Pressure-Low Allowable Value is selected high enough to detect imminent TCV fast closure.This protection is required consistent with the safety analysis whenever THERNAL POWER is)30%RTP.Below 30%RTP, the Reactor Vessel Steam Dome Pressure-High and the APRN Fixed Neutron Flux-High Functions of the RPS are adequate to maintain the necessary safety margins.ACTIONS A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable EOC-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable EOC-RPT instrumentation channel.A.l With one or more channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Actions B.1 and B.2 Bases), the EOC-RPT System is capable of performing the intended function.However, the reliabi-lity and redundancy (continued) BFN-UNIT 2 B 3.3-84 Amendment il~/Q~ik EOC-RPT Instrumentation B 3.3.4.1 ACTIONS A.1 (continued) of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function.Therefore, only a limited time is allowed to restore compl.iance with the LCO.Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of an EOC-RPT, 72 hours is provided to restore the inoperable channels (Required Action A.I)or apply the EOC-RPT inoperable NCPR limit.Alternately, the inoperable channels may be placed in trip (Required Action A.2)since this would conservatively compensate for the inoperability, restore capability to accommodate a single fai,lure, and allow operation to continue.As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the br'eaker may be inoperable such that it will not open).If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an RPT, or if the inoperable channel is the result of an inoperable breaker), Condition C must be entered and its Required Actions taken.B.l and 8.2 Required Actions B.I and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining EOC-RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped.Alternately, Required Action B.2 requires the NCPR limit for inoperable EOC-RPT, as specified in the COLR, to be applied.This also restores the margin to NCPR assumed in the safety analysis.The 2 hour Completion Time.is sufficient time for the operator to take corrective action, and takes into account (continued) BFN-UNIT 2 B 3.3-85 Amendment ili ili EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued) the likelihood of an event requiring actuation of the.EOC-RPT instrumentation during this period.It is also consistent with the 2 hour Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a NCPR violation. C.l With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to<30%RTP within 4 hours.The allowed Completion Time of 4 hours is reasonable, based on operating experience, to reduce THERMAL POWER to<30%RTP from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref.5)assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary. SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analysis of Reference 5.(continued) BFN-UNIT 2 B 3.3-86 Amendment 15~gg~ EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.4.1.2 This SR ensures that an EOC-RPT initiated from the, TSV-Clbsure and TCV Fast Closure, Trip.Oil Pressure-Low Functions will not be inadvertently bypassed when THERMAL POWER is w 30%RTP.This involves calibration of the bypass channels.Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 30%RTP, either due to open main turbine bypass valves or other reasons), the affected TSV-Closure and TCV Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can, be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met with the channel considered OPERABLE.The Frequency of 18 months is based on engineering judgment and reliability of the components. SR 3.3.4.1.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the, sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.SR 3.3.4.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.The system functional test of the pump breakers is included as a part of this test, overlapping the LOGIC SYSTEM FUNCTIONAL TEST, to provide complete testing of the associated safety function.Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would also be inoperable.(continued) BFN-UNIT 2 B 3.3-87 Amendment iSi ggi 0 EOC-RPT Instrumentation B 3.3.4.1 SURVEILLANCE RE(UIREMENTS SR 3.3.4.1.4 (continued) The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed'ith the reactor at power.Operating experience.has shown these components usually pass the Surveillance when.performed at the 18 month Frequency. REFERENCES 1.FSAR, Figure 7.9-2 (EOC-RPT logic diagram).2.FSAR, Section 7.9.4.5.3.FSAR,.Sections 14.5.1.1 and 14.5.1.3.4.FSAR,.Section 4.3.5.5.GENE-770-06-1,"Bases For Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times'For Selected Instrumentation Technical Specifications," February 1991.6.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-'UNIT 2 B 3.3-88 Amendment ili~i 0 ATWS-RPT Instrumentation B 3.3.4.2 B'3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should)occur, to lessen the effects of an ATWS event.Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level-Low or Reactor Steam Dome Pressure-High setpoint is reached, the recirculation pump motor breakers trip.The ATWS-RPT System (Ref.I)includes sensors, relays, bypass capability, circuit breakers, and switches that are.necessary to cause initiation of an RPT.The channels include electronic equipment (e.g., trip units)that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure-High and two channels of Reactor Vessel Water Level-Low in each.trip system.Each ATWS-RPT trip system is a two-out-of-two logic for each Function.Thus, either two Reactor Water Level-Low or two Reactor Pressure-High signals are needed to trip a trip system.The outputs.of the channels in.a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective motor breakers)' There are two motor breakers provided for each of the two recirculation pumps for a total of four breakers.The output of each trip system is provided to both recirculation pump breakers.BFN-UNIT 2 B 3.3-89 (continued) Amendment ili iS~45 ATWS-RPT Instrumentation B 3.3.4.2 BASES (continued) APPLICABLE SAFETY'ANALYSES, LCO, and APPLICABILITY The'ATWS-RPT is not assumed in the safety analysis.The ATWS-RPT initiates an RPT to aid in preserving the integrity of the fuel cladding following events in which a scram does not, but should, occur.Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4'f the NRC Policy Statement (Ref.3).The OPERABILITY of the ATWS-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.2.3. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). ATWS-RPT Channel OPERABILITY also includes the associated recirculation pump motor breakers.A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.Allowable Values are specified for each ATWS-RPT Function specified in the LCO.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints. are then determined accounting for the remaining instrument. errors (e.g., drift).The trip=setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and environmental effects are accounted for.(continued) BFN-UNIT 2 B 3.3-90 Amendment iki ik~0 ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) The individual Functions are required to be OPERABLE in MODE 1 to protect against catastrophic/multiple failures of the Reactor Protection System by providing a diverse trip to mitigate the consequences of a postulated ATWS event.The Reactor Steam Dome Pressure-High and Reactor Vessel Water Level-Low Functions are required to be OPERABLE in MODE 1, since the reactor is producing significant power and the recirculation system could be at high flow.During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.In MODE 2, the reactor is at low power and.the recirculation system is at low flow;thus, the potential is low for a pressure increase or low water level, assuming an ATWS event.Therefore, the ATWS-RPT is not necessary. In MODES 3 and 4, the reactor is shut.down with all control rods inserted;thus, an ATWS event is not significant and.the possibility of a significant pressure increase or low water level is negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant. In addition, the reactor pressure vessel (RPV)head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB)exists.The specific Applicable Safety Analyses and LCO discussions are listed below on a Function by Function basis.a.Reactor Vessel Water Level-Low Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel.The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.(continued) BFN-UNIT 2 8 3.3-91 Amendment ili il~ ATMS-RPT Instrumentation.B 3.3.4.2 APPLICABLE a.SAFETY ANALYSES, LCO, and APPLICABILITY Reactor Vessel Water Level-Low (continued) Four channels of Reactor Vessel Water Level-Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal.The Reactor Vessel Mater Level-Low Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation. b.Reactor Steam Dome Pressure-~Hi h Excessively high RPV pressure may rupture the RCPB.An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentia'lly result in fuel failure and overpressurization. The Reactor Steam Dome Pressure-High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety/relief valves, limits the peak RPV pressure to less than the ASNE Section III Code limits.The Reactor Steam Dome Pressure-High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.Four channels of Reactor Steam Dome Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid'ignal. The Reactor Steam Dome Pressure-High Allowable Value is chosen to provide an adequate margin to the ASNE Section III Code limits.ACTIONS A Note has been provided to modify the ACTIONS related to ATWS-RPT instrumentation channels.Section 1.3, Completion (continued) BFN-UNIT 2 8 3.3-92 Amendment il~i ib ATWS-RPT Instrumentation B 3.3.4.2 BASES ACTIONS (continued) Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition, However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel.A.l and A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained (refer to Required Actions..B. 1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function.However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function.Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status.Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1).Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open).If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT), or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its Required Actions taken.(continued) BFN-UNIT 2 B 3.3-93 Amendment ~~I~ ATWS-RPT Instrumentation B 3.3.4.2 ACTIONS (continued) B.l Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped.This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the recirculation pump motor breakers to be OPERABLE or in trip.The 72 hour Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of channels)and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining ATWS-RPT trip capability. C.1 Required Action C.1 is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function maintaining ATWS-RPT trip capability is discussed in the Bases for Required Action B.1 above.The 1 hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period.D.1 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply.To achieve this status, the plant must be brought to at least MODE 2 within 6 hours.The allowed Completion Time of 6 hours is.reasonable,'based on operating experience, both to reach MODE 2 from full power conditions and to remove a (continued) BFN-UNIT 2 B 3.3-94 Amendment ili ATMS-RPT Instrumentation B 3.3.4.2 BASES ACTIONS D.l (continued) recirculation pump from service in an orderly manner and without challenging plant systems.SURVEILLANCE 'REQUIREMENTS The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based'n the reliability analysis (Ref.2)assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary. SR 3.3.4.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly, between each CHANNEL CAL I BRAT ION.Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument 'has drifted outside its limit.(continued) BFN-UNIT 2 B 3.3-95 Amendment ili~~ ATWS-RPT Instrumentation B 3.3.4.2 BASES SURVEILLANCE REQUIREMENTS SR 3.3.4.2.1 (continued) The Frequency is based upon operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO., SR 3.3.4.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of'Reference 2.SR 3.3.4.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the.necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.SR 3.3.4.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channels The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function.Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would be inoperable.(continued) BFN-UNIT 2 B 3.3-96 Amendment ili 5QI ATWS-RPT Instrumentation B 3.3.4.2 SURVEILLANCE REQUIREMENTS SR 3.3.4.2.4.(continued) The 18 month Frequency is based on the need to perform this Surveillance under.the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown these components usually pass the Surveillance when performed. at the 18 month Frequency. REFERENCES 1.FSAR Section 7, 19.2.GENE-770-06-1,"Bases for Changes To Surveillance Test Intervals and Al.lowed Out-of-Service Times For Selected Instrumentation Technical Specifications,"'February 1991.3.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-97 Amendment Oi~i ECCS Instrumentation B 3.3.5.1 B 3.3 INSTRUMENTATION B 3.3.5.1 Emergency Core Cooling System (ECCS)Instrumentation BASES BACKGROUND The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient. For most anticipated operational occurrences and Design Basis Accidents (DBAs), a wide range, of dependent and independent parameters are monitored. The ECCS instrumentation actuates core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection (HPCI), Automatic Depressurization System (ADS), and the diesel generators (DGs).The equipment involved with each of these systems is described in the Bases for LCO 3.5.1,"ECCS-Operating." Core S ra S ste The CS System may be initiated by automatic means.Each pump can be controlled manually by a control.room remote switch.Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or both Drywell Pressure-High and Reactor Steam Dome Pressure-Low.Reactor water level and drywell pressure are monitored by four redundant transmitters, which are, in turn, connected to four trip units.The outputs of these trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic (i.e., two trip systems)for each Function.The Reactor Steam Dome Pressure-Low variable is monitored by two transmitters for each subsystem. The outputs from these transmitters are connected to relays arranged in a one-out-of-two logic.The high drywell pressure initiation signal is a sealed in signal and must be manually reset.Upon receipt of an initiation signal, if normal AC power is available, the four core spray pumps start one at a time, in order, at 0, 7, 14, and 21 seconds.If normal AC power is not available, (continued) BFN-UNIT 2 B 3.3-98 NENDMENT il~ik~0 ECCS Instrumentation B 3.3.

5.1 BACKGROUND

Core S ra S stem (continued) the four core spray pumps start seven seconds after standby power becomes available.(The LPCI pumps start as soon as standby power is available.) The CS test line isolation valve is closed on a CS initiation signal to allow full system flow assumed in the accident analyses.The CS pump discharge flow is monitored by a flow switch.When the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum flow return line valve is opened.The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the accident analysis.The CS System logic also receives, signals from transmitters which monitor the pressure in the reactor to ensure that, before the injection valves open, the reactor pressure has fallen to a value below the CS System's maximum design pressure.Reactor pressure is monitored by four redundant transmitters, which are, in turn, connected to four trip units (two per subsystem). The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two logic for each CS subsystem. Low Pressure Coolant In'ection S stem The LPCI is an operating mode of the Residual Heat Removal (RHR)System, with two LPCI subsystems. The LPCI subsystems may be initiated by automatic or manual means.Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or both Drywell Pressure-High and Reactor Steam Dome Pressure-Low.Each of these diverse variables is monitored by four redundant transmitters, which, in turn, are connected to four trip units.The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two-taken twice logic (i.e., two trip systems)for each Function.(continued) BFN-UNIT 2 B 3.3-99 Amendment ~i ik~il ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Low Pressure Coolant I ection S stem (continued) Once an initiation signal is received by the LPCI control circuitry, the signal is sealed in until manually reset.Upon receipt of an initiation signal, if normal AC powe}is available, the four RHR (LPCI)pumps start one at a time, in order,'t 0, 7, 14, and 21 seconds.If normal AC power is not available, the four pumps start simultaneously, with no delay, as soon as the standby power source is available. Each LPCI subsystem's discharge flow is monitored by a flow switch.When a pump is running and discharge flow is low enough so that pump overheating may occur, the respective minimum flow return line valve is opened.If flow is above the minimum flow setpoint, the valve is automatically closed.However, LPCI flow rates assumed in the LOCA analyses can be achieved with the minimum flow valve in the open position.The RHR test line suppression pool cooling isolation, valve, suppression pool spray isolation valves, and containment spray isolation valves (which are also PCIVs)are also closed on a LPCI initiation. signal to allow the full system flow assumed in the accident analyses and maintain primary containment isolated in the event LPCI is not operating. The LPCI System monitors the pressure in the reactor to ensure.that, before an injection valve opens, the reactor pressure has fallen to a value below the LPCI System's maximum design pressure.The variable is monitored by four redundant transmitters, which are, in turn, connected to multiple trip units.The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken.twice logic.Additionally, these instruments function to initiate closure of the recirculation pump discharge valves to ensure that LPCI flow does not bypass the core when it injects into the recirculation lines.Low reactor water level in the shroud is detected by two additional instruments which inhibit the manual initiation of other modes of RHR (e.g., suppression pool cooling)when (continued) BFN-UNIT 2 B 3.3-100 NENDHENT il~il~ ECCS Instrumentation B 3.3.

5.1 BACKGROUND

Low Pressure Cool nt In'ection S ste (continued) LPCI is required.Manual overrides for the inhibit logic are provided.i es ure Coola t ect o S ste The HPCI System may be initiated by either automatic or manual means.Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low, Level 2 or Drywell Pressure-High. Each of these variables is monitored by four redundant transmitters, which are, in turn, connected to multiple trip units.The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each Function.The HPCI pump discharge flow is monitored by a flow switch.Upon automatic initiation, when the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum, flow return line valve is opened.The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the accident analysis.The HPCI test line isolation valve is closed upon receipt of a HPCI initiation signal to allow the full system flow assumed in the accident analysis.The HPCI System also monitors the water levels in the HPCI pump supply header from the condensate storage tank (CST)and the suppression pool because these are the two sources of water for HPCI operation. Reactor grade water in the CST is the normal source.Upon receipt of a HPCI initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position)unless both suppression pool suction valves are open.If the water level in the HPCI pump supply header from the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes.Two level switches are used to detect low water level in the, HPCI pump supply header from the CST.Either switch can cause the suppression pool suction valves to open and the CST suction valve to close.(continued) BFN-UNIT 2 B 3.3-101 AMENDMENT il~ili ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND r s e Coo ant I ec'o S ste (continued) The suppression pool suction val.ves also automatically open and the CST suction valve closes if high water level is detected, in the suppression pool.To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.The HPCI provides makeup water to the reactor until the reactor vessel water level reaches the Reactor Vessel Water Level-High, Level 8 trip, at which time the HPCI turbine trips,.which causes the turbine's stop valve to close.The logic is two-out-of-two to provide high reliability of the HPCI System.The HPCI, System automatically restarts if a Reactor Vessel Mater Level-Low Low, Level 2 signal is subsequently received.tomatic De ressur zat on S ste The ADS may be initiated by either automatic or manual means.Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Low Low, Level 1;Drywell Pressure-High or ADS High Drywell Pressure Bypass Timer;confirmed Reactor Vessel Water Level-Low, Level 3;and CS or LPCI Pump Discharge Pressure-High are all present and the ADS Initiation Timer has timed out.There are two transmitters each for Reactor Vessel Water Level-Low Low Low, Level 1 and Drywell Pressure-High, and one transmitter for confirmed Reactor Vessel Water Level-Low, Level 3 in each of the two ADS trip systems.Each of these transmitters connects to a trip unit, which then drives a relay whose contacts form the initiation logic.Each ADS trip system includes a time delay between satisfying the initiation logic and the actuation of the ADS valves.The ADS Initiation Timer time delay setpoint chosen is long enough that the HPCI has sufficient operating time to recover to a level above Level 1, yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI fails to maintain that level.An alarm in the control room is annunciated when either of the ADS Initiation Timers is timing.Resetting the ADS initiation signals resets the ADS Initiation Timers.(continued) BFN-UNIT 2 B 3.3-102 NENDHENT il~ggi 0 ECCS Instrumentation B 3.3.

5.1 BACKGROUND

t ress o S ste (continued) The ADS also monitors the discharge pressures of the four LPCI pumps and the four CS pumps.Each ADS trip system includes two discharge pressure permissive switches from two of the four CS pumps (A and B for one trip system and C and D for the other trip system)and one discharge pressure permissive switch for each LPCI pump.The signals are used as a permissive for ADS actuation, indicating that there is a source of core coolant available once the ADS has depressurized the vessel.CS pumps (A or B and either C or D)or any one of the four LPCI pumps is sufficient to permit automatic depressurization. The ADS logic in each trip system is arranged in two strings.Each string has a contact from each of the following variables: Reactor Vessel Water Level-Low Low Low, Level 1;Drywell Pressure-High; or Low Water Level Actuation Timer.One of the two strings in each trip system must also have a confirmed Reactor Vessel Water Level-Low, Level 3.All contacts in both logic strings must close, the ADS initiation timer must time out, and a CS or LPCI pump discharge pressure signal must be present to initiate an ADS trip system.Either the A'r B trip system will cause all the ADS relief valves to open.Once the Drywell Pressure-High signal, the ADS High Drywell Pressure Bypass Timer, or the ADS initiation signal is present, it is individually sealed in until manually reset.Manual inhibit switches are provided in the control room for the ADS;however, their function.is not required for ADS OPERABILITY (provided ADS is not inhibited when required to be OPERABLE). Diesel Generators The DGs may be initiated by either automatic or manual means.Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or both Drywell Pressure-High and Reactor Steam Dome Pressure-Low. The DGs are also initiated upon loss of voltage signals.(Refer to the'Bases for LCO 3.3.8.1,"Loss of Power (LOP)Instrumentation," for a discussion of these signals.)Each of these diverse variables is monitored by four redundant transmitters, which are, in turn, connected to four trip (continued) BFN-UNIT 2 B 3.3-103 AMENDMENT ili ggi il ECCS Instrumentation B 3.3.

5.1 BACKGROUND

d ((dl units.The outputs of the four trip units are connected to relays whose contacts are connected to a one-out-of-two taken twice logic to initiate, all eight DGs{A, B, C, D, 3A, 3B, 3C, and 3D).The DGs receive their initiation signals from the CS System initiation logic.The DGs can also be started manually from the control room and locally from the associated DG room.The DG initiation signal is a sealed in signal and must be manually reset.The DG initiation logic is reset by resetting the associated ECCS initiation logic.Upon receipt of a loss of coolant accident (LOCA)initiation signal, each DG is automatically started, is ready to load in approximately 10 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open).The DGs will only energize their respective Engineered Safety Feature buses if a loss of offsite power occurs.(Refer to Bases for LCO 3.3.8.1.)mer e c E u'ent Coolin ater ECM S stem The EECW System, which distributes cooling water supplied by the RHR Service Water System pumps that are assigned as the principal supply to the EECW System (RHRSW pumps A3, B3, C3 and D3), may be initiated by automatic or manual means.Automatic initiation occurs for conditions of Reactor Vessel Mater Level-Low Low Low, Level 1 or Drywell Pressure-High with a Reactor Steam Dome Pressure-Low permissive. Each of these diverse variables is monitored by four redundant transmitters, which are, in turn, connected to four trip units.The EECM System receives its initiation signals from the DG initiation logic and the CS System initiation logic.The two RHRSW pumps (83 and D3)assigned to EECM and powered from shutdown boards in Units 1 and 2 will start automatically in less than 32.5 seconds after starting of a diesel generator or 30 seconds for a core spray pump in Units 1 and 2.The two RHRSW pumps (A3 and C3)assigned to EECW and powered from shutdown boards in Unit 3 will start automatically in less than 32.5 seconds after starting of a diesel generator or 30 seconds for a core spray pump in Unit 3.In addition, the signals that start the A3 and C3 pumps and the B3 and D3 pumps also start the Bl and Dl pumps and the Al and Cl pumps, respectively, when they are valved into the EECW header.BFN-UNIT 2 B 3.3-104 (continued) AMENDMENT il~ ECCS Instrumentation B 3.3.5.1 BASES (continued) APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The actions of the ECCS are explicitly assumed in the safety analyses of References 1, 2, and 3.The ECCS is initiated to preserve the integrity of the fuel cladding by limiting the post LOCA.peak cladding temperature to less than the 10 CFR 50.46 limits.ECCS instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.5).Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Each ECCS subsystem must also respond within its assumed response time.Table 3.3.5.1-1, footnote (b),, is added to show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation and actuation of other Technical Specifications (TS)equipment. Allowable Values are specified for each ECCS Function specified in the table.Nominal trip setpoints are specified in the setpoint cal'culations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel.water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived (continued) BFN-UNIT 2 B 3.3-105 NENDNENT ili ili ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG)initiation to mitigate the consequences of a design basis transient or accident.To ensure reliable ECCS and DG function, a combination of Functions is required to provide primary and secondary initiation signals.The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed.below on a Function by Function basis.Core S ra d Low Pressure Coo ant In'ectio S stems 2.a.Re cto Vesse Water eve-Low Low Low Level 1 Low reactor pressure vessel (RPV)water level indicates that the, capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.The low pressure ECCS, associated DGs, and EECW System are initiated at Level 1 to ensure that core spray and flooding functions are available to prevent or minimize fuel damage.The Reactor Vessel Water Level-Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 1 and 3.In addition, the Reactor Vessel Water Level-Low Low Low, Level 1 Function is directly assumed in the analysis of the recirculation line break (Ref..2).The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak'cladding temperature remains below the limits of 10 CFR 50.46.Reactor Vessel Water Level-Low Low Low, Level 1 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.(continued) BFN-UNIT 2 B 3.3-106 AMENDMENT iS~ ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Re cto Ve sel Water eve ow ow ow eve (continued) The Reactor Vessel Mater Level-Low Low Low, Level I Allowable Value is chosen to allow time, for the low pressure injection/spray subsystems to activate and provide adequate cooling.Four channels of Reactor Vessel Water Level-Low Low Low, Level I'unction are only required to be OPERABLE when the ECCS, DG(s), or EECW System are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS, DG, and EECM initiation. Refer to LCO 3.5.1 and LCO 3.5.2,-"ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; LCO 3.7.2,"Emergency Equipment Cooling (EECW)Systems and Ultimate Heat Sink (UHS)," for Applicability Bases for EECW System;and LCO 3.8.1,"AC Sources-Operating"; and LCO 3.8.2,"AC Sources-Shutdown," for Applicability Bases for the DGs.I b.b Dr e essure i High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB).The low pressure ECCS, associated DGs, and EECM System are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel damage.The Drywell Pressure-High Function, along with the Reactor Pressure-Low Function, is directly assumed in the analysis of the recirculation line break (Ref.2).The core cooling function of the ECCS, along.with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure.The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside.primary containment. The Drywell Pressure-High Function is required to be OPERABLE when the ECCS, DG, or EECM System are required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE.Thus, four channels of the CS and LPCI Drywell Pressure-High Function are required to be OPERABLE in.NODES I, 2, and 3'to ensure that no single instrument failure can preclude ECCS, DG, and (continued) BFN-UNIT 2 B 3.3-107 NENDMENT il~0 ECCS Instrumentation B 3.3'.5.1 BASES APPLICABLE SAFETY ANALYSES, L'CO, and APPLICABILITY e Pr ssure-(continued) EECW System initiation. In NODES 4 and 5, the Drywell Pressure-High Function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint.Refer to LCO 3.5.1 for Applicability Bases for the low pressure ECCS subsystems, LCO 3.7.2'for Applicability Bases for the EECW System, and to LCO 3.8.1 for Applicability Bases for the DGs..c.eactor S earn ome Pressu e-ow'ect on Permissive and ECCS I tiation Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems'aximum design pressure.The Reactor Steam Dome Pressure-Low is one of the Functions: assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3.In addition, the Reactor Steam Dome Pressure-Low Function is directly assumed in the analysis of the recirculation 'line break (Ref.2).The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.The Reactor Steam Dome Pressure-Low signals are initiated from fout pressure transmitters that sense the reactor dome pressure.The Allowable VaTue is low enough to prevent overpressurizing the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.Four channels of Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE when the ECCS is'required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.(continued) BFN-UNIT 2 B 3.3-108 NENDHENT ~i 45 ECCS Instrumentation 8 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) The minimum flow instruments are provided to protect the associated CS pumps from overheating when the pump is operating and the associated injection valve is not fully open.The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.The CS Pump Discharge Flow-Low Function is assumed to be OPERABLE and capable of closing the minimum flow valves to ensure that the CS flows assumed during the transients and accidents analyzed in References 1, 2, and 3 are met.The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.One flow switch per CS subsystem is used to detect the associated subsystems'low rates.The logic is arranged such that each flow switch causes its associated minimum flow valve to open.The logic will close the minimum flow valve once the closure setpoint is exceeded.The Pump Discharge Flow-Low Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough (based on engineering judgment)to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.Each channel of Pump Discharge Flow-Low Function (two CS channels)is only required to be OPERABLE when the associated ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude the ECCS function.Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems. l.e 2.f.Core S ra and Low Pressure Coolant In'ection~P<<-i Il l R1 The reaction of the low pressure ECCS pumps to an initiation signal depends on the availability of power.If normal power (offsite power)is not available, the four RHR (LPCI)pumps start simultaneously after the standby power source (four diesel generators) is available while the CS pumps start simultaneously after a seven-second time delay.This (continued) BFN-UNIT 2 B 3.3-109 AMENDMENT 0 ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY'NALYSES, LCO, and APPLICABILITY (continued) l.e 2.f.Core S ra and Low Pressure Coolant In'ection time delay.allows the start of LPCI pumps to avoid overloading the diesel generators. When normal power is available, the CS and RHR pump starts are staggered by shutdown board (i.e., A pumps start at 0 seconds, B pumps start at 7 seconds, C pumps start at 14 seconds, and 0 pumps start at 21 seconds).The purpose of this time delay, when power is being provided from the normal power source (offsite), is to stagger, the start of the CS and LPCI pumps, thus limiting the starting transients on the 4.16 kV shutdown buses.The CS and LPCI Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.There are four CS Pump and six LPCI Pump Start-Time Delay Relays when power is being provided from the normal power source, one in each of the pump start logic circuits (LPCI pumps C and D have two time delay relays).While each time delay relay is dedicated to a single pump start logic, a single failure of a CS or LPCI Pump Start-Time Delay Relay could result in the loss of normal power to a 4.16 kV shutdown board due to a voltage transient on the associated shutdown bus (e.g., as in the case where ECCS pumps on one shutdown bus start simultaneously due to an inoperable time delay relay).This would result in the affected board being powered by the associated diesel.Ther'efore, the worst case single failure would be failure of a single pump to start due to a relay failure leaving seven of the eight low pressure ECCS pumps OPERABLE;thus, the single failure criterion.is met (i..e., loss of one instrument does not preclude ECCS initiation). Since the CS pumps are 50%capacity pumps, the LOCA analysis does not take credit for a CS loop if one of the pumps is inoperable. Therefore, a 4.16 kV shutdown board failure results in the loss of one RHR pump and one CS loop (two CS pumps)for the LOCA analysis.The Allowable Value for the CS and LPCI Pump (continued) BFN-UNIT 2 B 3.3-110 AMENDMENT i il ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY l.e 2.f.Core S ra and Low Pressure Coolant In'ection Start-Time Delay Relays is chosen to be long enough so that most of the starting transient of the first set of pumps is complete before starting the second set of pumps on the same 4.16 kV shutdown bus and short enough so that ECCS operation is not degraded.There are also four CS and six LPCI Pump Start-Time Delay Relays when power is being provided by the standby source, one in each of the pump start logic circuits (LPCI pumps C and 0 have two time delay relays).While each relay is dedicated to a single pump start logic, a single failure of a Pump Start-Time Delay Relay could result in the failure of the two low pressure ECCS pumps (CS and LPCI)powered from the same shutdown board, to perform their intended function (e.g., as in the case where both ECCS pumps on one shutdown board start simultaneously due to an inoperable time delay relay).This still leaves six of eight low pressure ECCS pumps OPERABLE;thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). As stated above, since the LOCA analysis does not take credit for a CS loop if one of the pumps is inoperable, the loss of a 4.16 kV shutdown board effectively results in the loss of one LPCI pump and one CS loop (two CS pumps).The Allowable Value for the CS and LPCI Pump Start-Time Delay Relays is chosen to be long enough so that most of the starting transient for the LPCI pump is complete before starting the CS pump on the same 4.16 kV shutdown board and short enough so that ECCS operation is not degraded.Each CS and LPCI Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated CS and LPCI subsystems are required to be OPERABLE.Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the CS and LPCI subsystems. 2.d.Reactor Steam Dome Pressure-Low Recirculation Dischar e Valve Permissive Low, reactor steam dome pressure signals are used as permissives for recirculation discharge valve closure.This ensures that the LPCI subsystems inject into the proper RPV (continued) BFN-UNIT 2 B 3.3-111 ANENDHENT il ib ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 2.d.Reactor Steam Dome Pressure-Low Recirculation Dischar e Valve Permissive (continued) location assumed in the safety analysis.The Reactor Steam Dome Pressure-Low is one of the Functions assumed to be OPERABLE and capable of closing the valve during the transients analyzed in References 1 and 3.The core cooling function of the ECCS, along with the scram acti'on of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.The Reactor Steam Dome Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref.2).The Reactor Steam.Dome Pressure-Low signals are initiated from four pressure transmitters that sense the reactor dome pressure.The Allowable Value is chosen to ensure that the valves close prior to commencement of LPCI injection flow into the core, as assumed in the safety analysis.Four channel's of the Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE in NODES 1, 2, and 3 with the associated recirculation pump discharge valve open..With the valve(s)closed, the function of the instrumentation has been performed; thus, the Function is not required.In NODES 4 and 5, the loop injection location is not critical since LPCI injection through the recirculation loop in either direction will still ensure that LPCI flow reaches the core (i.e., there is no significant reactor steam dome back pressure). 2.e..Reactor Vessel Water Level-Level 0 The Level 0 Function is provided as a permissi,ve to allow the RHR System to be manually aligned from the LPCI mode to the suppression pool cooling/spray or drywell spray modes.The permissive ensures that water in the vessel is approximately two thirds core height before the manual transfer is allowed.This ensures that LPCI is available to prevent or.minimize fuel damage.This function may be overridden during accident conditions as allowed by plant procedures. Reactor Vessel Water Level-Level 0 Function is implicitly assumed in the analysis of the recirculation line break (Ref.2)since the analysis assumes that no LPCI flow diversion occurs when reactor water level is below Level 0.(continued) BFN-UNIT 2 B 3.3-112 AMENDMENT hli'l~i5 ECCS,Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.e.Reactor Vessel Water Level-Level 0 Reactor Vessel Water Level-Level 0 signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column.of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.The Reactor Vessel Water Level-Level 0 Allowable Value is chosen to allow the low pressure core flooding, systems to activate and provide adequate cooling before allowing a manual transfer.Two channels of the Reactor Vessel Water Level-Level 0 Function are only required to be OPERABLE in MODES 1, 2, and 3.In MODES 4 and 5, the specified initiation time of the LPCI subsystems is not assumed, and other administrative controls are adequate to control.the valves.that this Function isolates (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normally not used).HPCI S stem 3.a.Reactor Vessel Water Level-Low Low Level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, the HPCI System is initiated at'Level 2 to,maintain level above the top of the active fuel.The Reactor Vessel Water Level.-Low'Low, Level 2 is one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in References 1, 2, and 3.The core cool'ing function of the ECCS, al'ong with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below.the limits, of 10 CFR 50.46.Reactor Vessel Water Level-Low Low, Level, 2 signals are initiated'rom four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.The Reactor Vessel Water Level.-Low Low, Level 2 Allowable Value is high enough such.that for complete loss of (continued) BFN-UNIT 2 B 3.3'-113 AMENDMENT ili il~ ECCS Instrumentation B 3.3.5.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.a.Reactor Vessel Water Level-Low Low Level 2 (continued) feedwater flow, the Reactor Core Isolation Cooling (RCIC)System flow with HPCI assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Level-Low Low Low, Level l.Four channels of Reactor Vessel Water Level-Low Low,'Level 2 Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no.single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI.Applicability Bases.3.b.Dr well Pressure-Hicih High pressure in the drywell could indicate a break in the RCPB.The HPCI System is initiated upon receipt of the Drywell Pressure-High Funct'ion in order to minimize the possibility of-fuel damage.The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding, temperature remains below the-limits of 10 CFR 50.46.High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure.The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. Four channels of the Drywell Pressure-High Function are required to be OPERABLE when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI Applicability Bases.3..R V<<III I 1-~Hih High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel.Therefore, the Level 8 signal is used to trip the HPCI turbine to prevent overflow into the main steam lines (MSLs).The Reactor Vessel Water Level-High, Level 8 Function is not assumed in the accident and transient analyses.It was retained since it is a (continued) BFN-UNIT 2 B 3.3-114 AMENDMENT ili<g>>il ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.c.Reactor Vessel Water Level-Hi h Level 8 (continued) potentially significant contributor to risk, thus it meets Criterion 4 of the NRC Policy Statement (Ref.5).Reactor Vessel Water Level-High, Level 8 signals for HPCI are initiated from two level transmitters from the narrow range water level measurement instrumentation. The Reactor Vessel Water Level-High, Level 8 Allowable Value is chosen to prevent flow from the HPCI System from overflowing into the NSLs.Two channels of Reactor Vessel Water Level-High, Level 8 Function are required to be OPERABLE only when HPCI is required to be OPERABLE.Refer to LCO 3.5.1 for HPCI Applicability Bases.3.d.Condensate Header Level-Low Low level in.the CST indicates the unavailability of an adequate supply of makeup water from this normal source.Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST.However, if the water level in the HPCI pump supply header from the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes.This ensures that an adequate supply of makeup water is available to the HPCI pump.To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.The Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI)since the analyses assume that the HPCI suction source is the suppression pool.Condensate Header Level-Low signals are initiated from two level switches.The logic is arranged such that either level switch can cause the suppression pool suction valves to open and the CST suction valve to close.The Condensate Header Level-Low Function Allowable Value is high enough to ensure adequate pump suction head while water is being taken from the CST.0 (continued) BFN-UNIT 2 B 3.3-115 AMENDMENT ili<Qi il ECCS Instrumentation B 3.3.5.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.d.Condensate Header Level-Low (continued) One channel of the Condensate Header Level-Low Function is required to be OPERABLE only when HPCI is required to be OPERABLE.Refer to LCO 3.5al for HPCI Applicability Bases.S.e.So ressioo Pool Mater Level-~Hi ir Excessively high suppression pool water could result in the loads on the suppression pool exceeding design values should there be a blowdown of the reactor vessel pressure through the safety/relief valves.Therefore, signals indicating high suppression pool water level are used to transfer the suction source of HPCI from the CST to the suppression pool to e1iminate the possibility of HPCI continuing to provide additional water from a source outside containment. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.This Function is impl.icitly assumed in the accident and transient analyses (which take credit for HPCI)since the analyses assume that the HPCI suction source is the suppression pool.Suppression 'Pool Water Level-High signals are initiated from two level switches.The logic is arranged such that either switch can cause the suppression pool suction valves to open and the CST suction valve to close.The Allowable Value for the Suppression Pool Water Level-High Function is chosen to ensure that HPCI will be aligned for suction from the suppression pool before the water level reaches the point at which suppression pool design loads would be exceeded.One channel of Suppression Pool Water Level-High Function is required to be OPERABLE only when HPCI is required to be OPERABLE.Refer to LCO 3.5.1 for HPCI Applicability Bases.3.f.Hi h Pressure Coolant In'ection Pum Dischar e Fl-~LB The minimum flow instruments are provided to protect the HPCI pump from overheating when the pump is operating at reduced flow.The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed (continued) BFN-UNIT 2 B 3.3-116 AMENDMENT O~il~ ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.f.Hi h Pressure Coolant In'ection Pum Dischar e when the flow rate is adequate to protect the pump.The High Pressure Coolant Injection Pump Discharge Flow-Low Function is assumed to be OPERABLE and capable of closing the minimum flow valve to ensure that the ECCS flow assumed during the transients and accidents analyzed in References 2 and 3 are met.The core cooling function of the ECCS, along with the scram action.of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.One flow transmitter is used to detect the HPCI System's flow rate.The logic is arranged such that the transmitter causes the minimum flow valve to open.The logic will close the minimum flow valve once the closure setpoint is exceeded.The High Pressure Coolant Injection Pump Discharge Flow-Low Allowable Value is.high enough to ensure that pump flow rate is sufficient to protect the pump, yet low enough (based on engineering judgment)to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.~One channel is required to be OPERABLE when the HPCI is required to be OPERABLE.Refer to LCO 3.5.1'or HPCI Applicability Bases.Automatic De ressurization S stem 4.a 5.a.Reactor Vessel Water Level-Low Low Low Level 1 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, ADS receives one of the signals necessary for initiation from this Function.The Reactor Vessel Water Level-Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ADS during the accident analyzed in Reference 2.The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.(continued) BFN-UNIT 2 B 3.3-117 AMENDMENT il~ili il ECCS Instrumentation B 3.3.5.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 4.a 5.a.Reactor Vessel Water Level-Low Low Low Level 1 (continued) Reactor Vessel Water Level-Low Low Low, Level 1 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-Low Low Low, Level 1 Function are required to be OPERABLE only when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B.Refer to LCO 3.5.1 for ADS Applicability Bases.The Reactor Vessel Water Level-Low Low Low, Level 1 Allowable Value is chosen to allow time for the low pressure core flooding systems to initiate and provide adequate cooling.4.b 5.b.Dr well Pressure-~Hi h High pressure in the drywell could indicate a break in the RCPB.Therefore, ADS receives one of the signals necessary for initiation from this Function in order to minimize the possibility of fuel damage.The Drywell Pressure-High is assumed to be OPERABLE and capable of initiating the ADS during the accidents analyzed in Reference 2.The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.Drywell-Pre'ssure-High signals are initiated from four pressure transmitters that sense drywell pressure.The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment. Four channels of Drywell Pressure-High Function are only required to be OPERABLE when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B.Refer to LCO 3.5.1 for ADS Applicability Bases.(continued) BFN-UNIT 2 B 3.3-118 AMENDMENT i ik~, ECCS Instrumentation 8 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)- 4.c 5.c.Automatic De ressurization S stem Initiation Timer The purpose of the Automatic Depressurization System Initiation Timer is to delay depressurization of the reactor vessel to allow the HPCI System time to maintain reactor vessel water level.Since the rapid depressurization caused by ADS operation is one of the most severe transients on the reactor vessel, its occurrence should be limited.By delaying initiation of the ADS Function, the operator is given the chance to monitor the success or failure of the HPCI System to maintain water level,@'and then to decide whether or not to allow ADS to initiate, to delay initiation further by recycling the timer, or to inhibit initiation permanently. The Automatic Depressurization System Initiation Timer, Function, is assumed to be OPERABLE for the accident analyses of Reference 2 that require ECCS initiation. There are two Automatic Depressurization System.Initiation Timer relays, one in each of the two AOS trip systems.The Allowable Value for the Automatic Depressurization System Initiation Timer is chosen so that there is still time after'epressurization for the low pressure ECCS subsystems to provide adequate core cooling.Two channels of the Automatic Depressurization System Initiation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation.(One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B.Refer to LCO 3.5.1 for ADS Applicability Bases.4.d 5.d.Reactor Vessel Water Level-Low Level 3 Confirmator The Reactor Vessel Water Level-Low, Level 3 (Confirmatory) Function is used by the ADS only as a confirmatory low water level signal.ADS receives one of the signals necessary for initiation from Reactor Vessel Water Level-Low Low Low, Level 1 signals.In order to prevent spurious initiation of the ADS due to spurious Level 1 signals, a Level 3 (Confirmatory) signal must also be received before ADS initiation commences.(continued) BFN-UNIT 2 B 3.3-119 AMENDMENT ~ili 0 ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 4.d 5.d.Reactor Vessel Water Level-Low Level 3 dddTY dddLY d,~i d d d)LCO, and APPLICABILITY 'Reactor Vessel Water Level-Low, Level 3 (Confirmatory) signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variab1e leg)in the vessel.Two channels of Reactor Vessel Water Level-Low, Level 3 (Confirmatory) Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B.Refer to LCO 3.5.1 for ADS Applicability Bases.4.e 4.f 5.e 5.f.Core S ra and Low Pressure Coolant In ection Pum Dischar e Pressure-~Hi h The Pump Discharge Pressure-High signals from the CS and LPCI pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling water available once the ADS has depressurized the vessel.Pump Discharge Pressure-High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in Reference 2 with an assumed HPCI failure.For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the core cooling functions. This core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.Pump discharge pressure signals are initiated from twelve pressure transmitters, two on the discharge side of each RHR (LPCI)pump and one on the discharge side of each CS pump.There are two ADS low pressure ECCS pump permissives in each trip system.Each of these permissives receives inputs from all four RHR (L'PCI)pumps (different signals for each permissive) and, two CS pumps, two for each subsystem (different pumps for each permissive). In order to generate an ADS permissive in one trip system, it is necessary that (continued) BFN-UNIT 2 8 3.3-120 AMENDMENT /gi ill ik~ ECCS Instrumentation B 3.3.5.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 4.e 4.f 5.e S.f.Core S ra and Low Pressure Coolant In ection Pun Dischar e Pressure-~Hi h (continued) only one LPCI pump or two CS pumps (CS pumps A or B and either C or D)indicate the high discharge pressure condition. The Pump Discharge Pressure-High Allowable Value is less than the pump discharge pressure when the pump is operating in a full flow mode and high enough to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running.The actual operating point of this function is not assumed in any transient or accident analysis.However, this function is indirectly assumed to operate (in Reference 2)to provide the ADS permissive to depressurize the RCS to allow the ECCS low pressure systems to operate.Twelve channels of Core Spray and Low Pressure Coolant Injection Pump Discharge Pressure-High Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Four CS channels associated with CS pumps A through D and eight LPCI channels associated with LPCI pumps A through D are required for trip systems.Refer to LCO 3.5.1 for ADS Applicability Bases.4.5..Automatic De ressurization S stem Hi h Dr well Pressure B ass Timer One of the signals required for ADS initiation is Drywell Pressure-High.However, if the event requiring ADS initiation occurs outside the drywell (e.g., main steam line break outside containment), a.high drywell pressure signal may never be present.Therefore, the Automatic Depressurization System High Drywell Pressure Bypass Timer is used to bypass the Drywell Pressure-High Function after a certain time period has elapsed.Operation of the Automatic Depressurization System High Drywell Pressure Bypass Timer Function is not assumed in any accident analysis.The instrumentation was installed to meet requirements of NUREG-0737, Item II.K.3.18 (Ref.6)and is retained in the TS because ADS is part of the primary success path for mitigation of a DBA.(continued) BFN-UNIT 2 B 3.3-121 AMENDMENT

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 4.5..Automatic De ressurization S stem Hi h Dr e 1 Pressure 8 ass Timer (continued) There are two Automatic Depressurization System High Drywell Pressure Bypass Timer relays, one in each of the two ADS trip systems.The Allowable Value for the Automatic Depressurization System High Drywell Pressure Bypass Timer is chosen to ensure that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling.Two channels of the Automatic Depressurization System High Drywell Pressure Bypass Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Refer to LCO 3.5.1 for ADS Applicability Bases.ACTIONS A Note.has been provided to modify the ACTIONS related to ECCS instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel.A.1 Required Action A.l directs entry into the appropriate Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the table is Function dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.(continued) BFN-UNIT 2 B 3.3-122 AMENDMENT 4'i ECCS Instrumentation 8 3.3.5.1 BASES ACTIONS (continued) B.1 B.2 and.8.3 Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature(s).. Required Action B.1 features would be those that are initiated by Functions l.a, l.b, l.c, 2.a, 2.b, and 2.c (e.g., low pressure ECCS).The Required Action 8.2 system would be HPCI.For Required Action B.1, redundant automatic initiation capability is lost if (a)two or more Function l.a channels are inoperable and untripped such that both trip systems lose initiation capability, (b)two or more Function 2.a channels are inoperable and untripped such that both trip systems lose initiation capability, (c)two or more Function l.b channels are inoperable and untripped such that both trip systems lose initiation capability, (d)two or more Function 2.b channels are inoperable and untripped such that both trip systems lose initiation capability, (e)two or more Function l.c channels are inoperable and untripped such that both trip systems lose initiation capability, or (f)two or more Function 2.c channels are inoperable and untripped such that both trip systems lose initiation capability. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS, DGs, and EECW to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS, DGs, and EECW being concurrently declared inoperable. For Required Action B.2, redundant.automatic HPCI initiation capability is lost if two or more Function 3.a or two or more Function 3.b channels are inoperable and untripped such that the trip system loses initiation capability. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action B.3 is not appropriate and the HPCI System must be declared inoperable within 1 hour.As noted (Note 1 to Required Action B.1), Required Action B.1 is only applicable in NODES 1, 2, and 3.In NODES 4 and 5, the specific (continued) BFN-UNIT 2 8 3.3-123 AMENDMENT ik~0 ECCS Instrumentation B 3.3.5.1 ACTIONS B.l B.2 and B.3 (continued) initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower.Thus, a total loss of initiation capability for 24 hours (as allowed by Required Action B.3)is allowed during MODES 4 and 5.There is no similar Note provided for Required Action B.2 since HPCI instrumentation is not required in MODES 4 and 5;thus, a Note is not necessary. Notes are also provided (Note 2 to Required Action B.1 and the Note to Required Action 8.2)to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is performed. Required Action B.1 (the Required Action for certain inoperable channels in the low pressure.ECCS subsystems) is not applicable to Function 2.e, since this Function provides backup to administrative controls ensuring that operators do not divert LPCI flow from injecting into the core when needed.Thus, a total loss of Function 2.e capability for 24 hours is allowed, since the LPCI subsystems remain capable of performing their intended function.The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action B.l, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above.For Required Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated. Function in the same trip.system.The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.Because of the diversity of sensors available to provide initiation signals, and the redundancy of the ECCS design,,an allowable out of service time of 24 hours has been shown to be acceptable (Ref.4)to permit restoration of any (continued) BFN-UNIT 2 B 3.3'-124 AMENDMENT il 0 ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1 B.2 and B.3 (continued) inoperable channel to OPERABLE status.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.3.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.C.1 and C.2 Required Action C..l is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function result in redundant automatic initiation capability being lost for the feature(s). Required Action C.1 features would be those that are initiated by Functions l.e, 2.d, and 2.f (i.e., low pressure ECCS).Redundant automatic initiation capability is lost if either (a)two or more Function l.e channels are inoperable affecting CS pumps in different subsystems, (b)two or more Function 2.d channels are inoperable in the same trip system such that the trip system loses initiation capability, or (c)two or more or more Function 2.f channels are inoperable affecting two LPCI pumps.In this situation (loss of redundant automatic initiation capabil.ity), the 24 hour allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour.Since each inoperable channel would have Required Action C.l applied separately (refer to ACTIONS Note), each.inoperable channel would only require the affected portion of the, associated system.to, be.declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g., both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable. For Functions I.e, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps.As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3.(continued) BFN-UNIT 2.8 3.3-125 AMENDMENT O~0 Cl ECCS Instrumentation B 3.3.5.1 ACTIONS d.d dd d)In NODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower.Thus, a total loss of automatic initiation capability for 24 hours (as allowed by Required Action C.2)is allowed during HODES 4 and 5.Note 2 states that Required Action C.l is only applicable for Functions l.e, 2.d,'nd 2.f.Required Action C.l is also not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic).This loss was considered during the development of'Reference 4 and considered acceptable for the 24 hours allowed by Required Action C.2.The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the, normal"time zero" for beginning the allowed outage time"clock." For Required Action C.l, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above.The 1 hour Completion Time from discovery'f loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref.4)to permit restoration of any inoperable channel to OPERABLE status.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken.The Required Actions do not allow placing.the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events.(continued) BFN-UNIT 2 B 3.3-126 AMENDMENT 0 ECCS Instrumentation 8 3.3.5.1 BASES ACTIONS (continued) D.1 Required Action 0.1 is intended to ensure that appropriate actions are taken if an inoperable, untripped channel within the same Function results in a complete loss of automatic component initiation capability for the HPCI System.Since Table 3.3.5.1-1 only requires one channel to be OPERABLE, automatic component initiation capability is lost if the one required Function 3.d channel or the one required Function 3.e channel is inoperable and untripped. In this situation (loss of automatic suction swap), the HPCI system must be declared inoperable within 1 hour.As noted, Required Action D.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the Function is already performed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action D.1, the Completion Time only begins upon discovery that the HPCI System cannot be automatically aligned to the suppression pool due to an inoperable, untripped channels in the same Function.The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Core Spray Pump Discharge Flow-Low Bypass Function results in redundant automatic initiation capability being lost for the feature(s). For Required Action E.1, the features would be those that are initiated by Function 1.d (i.e., CS).Redundant automatic initiation capability is lost if two Function 1.d channels are inoperable. In this situation (loss of minimum flow capability), the 7 day allowance of Required Action E.2 is not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour.As noted (Note 1 to Required Action E.1), Required Action E.1 is only applicable in MODES 1, 2, and 3.In MODES 4 and 5, the (continued) BFN-UNIT 2 8 3.3-127 AMENDMENT 0 0 0 ECCS Instrumentation B 3.3.5.1 ACTIONS E.l and E.2 (continued) specific initiation time of the ECCS is not assumed and the probability of, a LOCA is lower.Thus, a total loss of initiation capability for 7 days (as allowed by Required Action E.2)is allowed during NODES 4 and 5.A Note is also provided (Note 2 to Required Action E.1)to delineate that Required Action E.1 is only applicable to low pressure ECCS Functions. Required Action E.1 is not applicable to HPCI Function 3.f since the loss of one channel results in a loss of the Function (one-out-of-one logic).This loss was considered during the development of Reference 4 and considered acceptable for the 7 days allowed by Required Action E.2.The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action E.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above.The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes.risk while allowing time for restoration of channels.If the instrumentation that controls the CS pump minimum flow valve is inoperable, such that the valve will not automatically open, extended CS pump operation with no injection path available could lead to pump overheating and fai,lure.If there were a failure of the instrumentation, such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor vessel injection path, causing insufficient core cooling.These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump protection and required flow.Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to,occur.The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low (continued) BFN-UNIT 2 B 3.3-128 AMENDMENT ~i 0 ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued) probability of a DBA occurring during the allowed out of service time..If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken.The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel-in al.l events.F.l and F.2 Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS.Redundant automatic initiation capability is lost if either (a)one or more Function 4.a channels and one or more Function 5.a channels are inoperable and untripped, (b)one or more Function 4.b channels and one or more Function S.b channels are inoperable and untripped, or (c)one or more Function 4.d channels and one or more Function 5.d channels are inoperable and untripped. In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate and all ADS valves must be declared inoperable within 1 hour after.discovery of loss of ADS initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action F.l, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above.The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an (continued) BFN-UNIT 2 B 3.3-129 AMENDMENT il~il ECCS Instrumentation B 3.'3.5.1 BASES ACTIONS~l d.(i d)allowable out of service time of 8 days has been shown to be acceptable (Ref.4)to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE.. If either HPCI or RCIC is inoperable, the time is shortened to 96 hours.If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days.If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the"time zero" for beginning the 8 day"clock" begins upon discovery of the inoperable, untripped channel.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action F.2.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.G.l and G.2 Required Action G.l is intended to ensure that appropriate actions are taken if multiple, inoperable channels within-similar ADS trip system Functions result in automatic initiation capability being lost.for the ADS.Automatic initiation capability is lost if either (a)one Function 4.c channel and one Function 5.c channel are inoperable, (b)a combination of Function 4.e, 4.f, 5.e, and.5.f channels are inoperable such that channels associated with five or more low pressure ECCS pumps are inoperable, or (c)one or more Function 4.g channels and one or more Function 5.g channels are inoperable. In this situation (loss of automatic initiation capability), the'96 hour or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within.1 hour after discovery of loss of ADS initiation capability.(continued) BFN-UNIT 2 B 3.3-130 AMENDMENT il~0 4l ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.l and G.2 (continued) The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action G.l, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above.The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref.4)to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action G.2).If either HPCI or RCIC is inoperable, the time shortens to 96 hours.If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the.96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable channel cannot exceed 8 days.If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the"time zero" for beginning the 8 day"clock" begins upon discovery of the inoperable channel.If the inoperable channel cannot be restored.to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken.The Required Actions do not allow placing the channel in.trip since this action would not necessarily result in a safe state for the channel in all events.With any Required Action and associated Completion Time not met, the associated feature(s) may be incapable of performing the.intended function, and the supported feature(s) associated with inoperable untripped channels must be declared inoperable immediately.(continued) BFN-UNIT 2 B 3.3-131 AMENDMENT 4S~il 41 ECCS Instrumentation.B 3.3.5.1 BASES SURVEILLANCE RE(UIREHENTS As noted in the beginning of the SRs, the SRs for each ECCS instrumentation Function are found in the SRs column of Table 3.3.5.1-1. The Surveillances are modified by a second Note (Note 2)to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a)for Functions 3.c and 3.f;and (b)for Functions other than 3.c and 3.f provided the associated Function or redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref..4)assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary. SR 3.3.5.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency is based upon operating experience that demonstrates channel failure is rare.The CHANNEL CHECK (continued) BFN-UNIT 2 B 3.3-132 NENDNENT !l~0 0 ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.5.1.1 (continued) supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.5.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on.the reliability analyses of Reference 4.SR 3.3.5.1.3 SR 3.3.5.1.4 and SR 3.3.5.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequencies of SR 3.3.5.1.3, SR 3.3.5.1.4, and SR 3.3.5.1.5 are based upon the magnitude of equipment drift in the setpoint analysis.SR 3.3.5.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel.The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.7.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety, function.(continued) BFN-UNIT,2 B 3.3-133 AMENDMENT ib ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS J55 3.3.5:1.6 (ti d)The 18 month Frequency is'based on the need to perform this Surveillance under, the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the, reactor at.power.Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. REFERENCES 1.FSAR, Section'8.5.2.FSAR, Section 6.5.3.FSAR, Chapter 14.4..NEDC-30936-P-A,"BWR Owners'roup Technical, Specification Improvement Analyses for-ECCS Actuation Instrumentation, Part 2," December 1988.5.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.E 6.NUREG-0737,"Clarification of TMI Action Plan Requirements," October 31, 1980.BFN-UNIT.2 B 3.3-134 AMENDMENT 0' RCIC System Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3.5.2 Reactor Core Isolation Cooling (RCIC)System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is unavailable, such that initiation of the low pressure Emergency Core Cooling Systems (ECCS)pumps does not occur.A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3,"RCIC System." The RCIC System may be initiated by either automatic or manual means.Automatic initiation occurs for conditions of reactor vessel Low Low water level.The variable is monitored by four transmitters that are connected to four trip units.The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement. Once initiated, the RCIC logic seals in and can be reset by the operator only when the reactor vessel water level signals have cleared.The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow.There are two sources of water for RCIC operation. Reactor grade water in the CST is the normal source and the suppression pool is the alternate source.Although the RCIC System does not monitor the water levels in the High Pressure Coolant Injection (HPCI)supply header from the condensate storage tank (CST)and the suppression pool, administrative controls are in place that direct the transfer from the CST to the suppression pool when the HPCI System automatically transfers on low HPCI.pump supply header level or high suppression pool level.The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8)trip (two-out-of-two logic), at whi'ch time the RCIC steam supply closes and the minimum flow valve closes, if open.The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2).BFN-UNIT 2 B 3.3-135 (continued) Amendment O~ib 4b RCIC System Instrumentation 8 3.3.5.2 BASES (continued) APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The function of the RCIC System to provide makeup coolant to the reactor is used to respond to transient events.The RCIC System is not an Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation meets Criterion 4 of the NRC Policy Statement (Ref.2).Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.2-1. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. A channel is.inoperable if its actual trip setpoint is not within its required Allowable Value.The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Allowable Values are.specified for each RCIC System instrumentation Function specified in'the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified accounts for instrument uncertainties appropriate to the Function.These uncertainties are described in the setpoint methodology. The individual Functions are required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure>150 psig since this is when RCIC is required to be OPERABLE.(Refer to LCO 3.5.3 for Applicability Bases for the RCIC System.)The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.(continued) BFN-UNIT 2 B 3.3-136 AMENDMENT il~ll RCIC System Instrumentation B 3.3.5.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 1.Reactor Vessel Water Level-Low Low Level 2 Low reactor pressure vessel (RPV)water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, the RCIC System is initiated at Level 2 to.assist in maintaining water level above the top of the active fuel.Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg),and'he pressure due to the actual water level (variable leg)in the vessel.The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level'l.Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.2.2 2 182 I I~IWR I 18 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel.Therefore, the Level 8 signal is used to close the RCIC steam supply valve to prevent overflow into the main steam lines (HSLs).Reactor Vessel Mater Level-High, Level 8 signals for RCIC are initiated from two level transmitters from the narrow range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.(continued) BFN-UNIT 2 B 3.3-137 Amendment i~0 RCIC System Instrumentation 'B 3.3.5.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 2.Reactor Vessel Water Level-Hi h Level 8 (continued) The Reactor Vessel Water Level-High, Level 8 Allowable Value is high enough to preclude closing the RCIC steam supply valve, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.Two channels of Reactor Vessel Water Level-High, Level 8 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE.Refer to LCO 3.5.3 for RCIC Applicability Bases.ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. B.l and B.2 Required Action B.l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss (continued) BFN-UNIT 2 B 3.3-138 NENDMENT ~i~' RCIC System Instrumentation B 3.3.5.2 ACTIONS B.l and 8.2 (continued) of automatic initiation capability for the RCIC System.In this situation'(loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour after discovery of loss of RCIC initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal"time zero" for beginning the allowed outage time"clock." For Required Action B.l, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two or more inoperable, untripped Reactor Vessel Water Level-Low Low, Level'2 channels such that the trip system loses initiation capability. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref.1)to permit restoration of any inoperable channel to OPERABLE status.For conservatism, in some transient analyses, RCIC flow rates were used rather than HPCI flow rates.If the inoperable, channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition D must be entered and its Required Action taken.C.l A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref.1)is (continued) BFN-UNIT 2 B 3.3-139 AMENDMENT

RCIC System Instrumentation B 3.3.5.2 BASES ACTIONS C.1 (continued) acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.l).A Required Action (similar.to Required Action B.l)limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required.This Condition applies to the Reactor Vessel Water Level-High, Level 8 Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. The Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events.D.1 With any Required Action and associated Completion Time not , met, the RCIC System may be incapable of performing the intended'unction, and the RCIC System must be declared inoperable immediately. SURVEILLANCE RE(UIREMENTS As noted in the beginning of the SRs, the SRs for each RCIC System instrumentation Function are found in the SRs column of'Table 3.3.5.2-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a)for up to 6 hours for Function 2;and (b)for up to 6 hours for Function 1, provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliability analysis (Ref.I)assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.(continued) BFN-UNIT 2 B 3.3-140 AMENDMENT il~0 RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.5.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more, serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency is based upon operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 1.(continued) BFN-UNIT 2 B 3.3-141 AMENDMENT 0 0 0 RCIC System Instrumentation B 3.3.5.2 SURVE ILL'ANCE REQUIREMENTS (cont'inued) SR 3.3.5'.2.3 A, CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency of SR 3.3.5.2.3 is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of.equipment drift in the setpoint analysis.'S 3.3.5.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel..The system functional testing performed in LCO 3.5.3 overlaps this Surveillance. to provide complete testing of the, safety function.The 18 month Frequency is, based on the, need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with.the reactor at power.Operating experience.has shown that these components usually pass the Surveillance when performed. at the 18 month Frequency. REFERENCES 1.GENE-770-06-2,."Addendum to Bases for Changes to Surveillance Test Intervals and Allowed.Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.2.NRC No.93-102,"Final Policy Statement on Technical Specifi'cation Improvements," July'3, 1993.BFN-UNIT 2 B 3.3-142 NENDMENT 0 0 Primary Containment Isolation Instrumentation B 3.3.6.1 8 3.3 INSTRUMENTATION B 3.3.6.1 Primary Containment Isolation Instrumentation BASES BACKGROUND The primary containment isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs).The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs).Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA.The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB)isolation. Host channels include electronic equipment (e.g., trip units)that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic.Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logics are (a)reactor vessel water level, (b)area ambient temperatures, (c)main steam line (HSL)flow measurement, (d)Standby Liquid Control (SLC)System initiation, (e)main steam line pressure, (f)high pressure coolant injection (HPCI)and reactor core isolation cooling (RCIC)steam line flow, (g)drywell pressure, (h)HPCI and RCIC steam line pressure, (i)HPCI and RCIC turbine exhaust diaphragm pressure, and (j reactor steam dome pressure.Redundant sensor input signals from each: parameter are provided for initiation of isolation. The only exception is SLC System initiation. Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below.BFN-UNIT.2 B 3.3-143 (continued) AHENDMENT 0 il Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND (continued) 1.Main Steam Line Isolation Most MSL Isolation Functions receive inputs from four channels.The outputs from these channels initiate isolation of the Group 1 isolation valves (main steam isolation valves (MSIVs)and MSL drain valves).The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the MSIVs.The outputs from the same channels are arranged into two two-out-of-two logic trip systems to isolate all MSL drain valves.The outputs from the Reactor Vessel Water Level-Low Low Low, Level 1 Function channels are arranged into two two-out-of-two logic trip systems to isolate the recirculation loop sample line valves.The exceptions to this arrangement are the Main Steam Line Flow-High Function and Area Temperature Functions. The Main Steam Line Flow-High Function uses 16 flow channels, four for each steam line.One channel from each steam line inputs to each of the four trip strings.Two trip strings make up each trip system and both trip systems must trip to cause an MSL'isolation. Each trip string has four inputs (one per MSL), any one of which will trip the trip string.The trip strings are arranged in a one-out-of-two taken twice.logic.The Main Steam Line'Space Temperature -High Function receives input from 16 channels.The logic is arranged similar to the Main Steam Line Flow-High Function.MSL Isolation Functions isolate the Group 1 valves.2.Primar Containment Isolation Most Primary Containment Isolation Functions receive inputs from four channels.The outputs from these channels are arranged into a one-out-of-two taken twice logic trip system.The isolation signal from this logic system isolates both containment isolation valves on a penetration. Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low, Level 3 Functions isolate the Group 2, 6 and 8 valves.The Reactor Vessel Water Level-Low, Level 3 Function also isolates Group 3 valves.(continued) BFN-UNIT 2 B 3.3-144 Amendment 0 il Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND (continued) 3 4.Hi h Pressure Coolant In'ection S stem Isolation and Reactor Core Isolation Coolin S stem Isolation Host Functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a one-out-of-two taken twice logic.Each of the two trip systems in each isolation group is connected to each of the two valves on each associated penetration. The exceptions are the HPCI and RCIC Steam Line Flow-High Functions. There are two channels for this Function which provide an isolation signal to both trip systems using one-out-of-two logic.Each of the two trip systems isolate both valves in an associated penetration. HPCI and RCIC Functions isolate the Group 4 and,5 valves.5.Reactor Water Cleanu S stem Isolation The Reactor Vessel Water Level-Low, Level 3 Isolation Function receives input from four reactor vessel water level channels.The outputs from the reactor vessel water level channels are connected into one-out-of-two taken twice trip systems.The SLC System Initiation Function provides an isolation signal to close both RWCU isolation valves.The Area Temperature -High Function receives input from twenty-four temperature monitors.There are four temperature sensors in each of the six areas where the RWCU piping and equipment are located.The four sensors in each area provide isolation signals to close both RWCU isolation valves using one-out-of-two logic.RWCU Functions isolate the Group 3 valves.The Reactor Vessel Water-Level-Low, Level 3 Function also isolates Group 2, 3, 6 and 8 valves.6.Shutdown Coolin S stem Isolation The Reactor Steam Dome Pressure-High Function receives input from two channels which provide one-out-of-two isolation logic to each isolation valve.The Shutdown Cooling System Isolation Functions isolate the Group 2 RHR Shutdown Cooling (SDC)valves.BFN-UNIT 2 B 3.3-145 (continued) Amendment il il 0 Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The isolation signals generated by the primary containment isolation instrumentation are implicitly assumed in the safety analyses of References 2 and 8 to initiate closure of valves to limit offsite doses.Refer to LCO 3.6.1.3,"Primary Containment Isolation Valves (PCIVs)," Applicable Safety Analyses Bases for more detail of the safety analyses.Primary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.7).Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the primary containment instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.'3.6.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Each channel must also respond within its assumed response time, where appropriate. Allowable Values are specified for each Primary Containment Isolation Function specified in the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal tr ip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation, (continued) BFN-UNIT 2 B 3.3-146 NENDNENT il~ Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY (continued) unce}tainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.Certain Emergency Core Cooling Systems (ECCS)and RCIC valves (e.g., minimum flow)also serve the dual function of automatic PCIVs.The signals that isolate these valves are also associated with the automatic initiation of the ECCS and RCIC.The instrumentation requirements and ACTIONS associated with these signals are addressed in LCO 3.3.5.1,"Emergency Core Cooling Systems (ECCS)Instrumentation," and LCO 3.3.5.2,"Reactor Core Isolation Cooling (RCIC)System Instrumentation," and are not included in this LCO.In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LCO 3.6.1.1,"Primary Containment." Functions that have different Applicabilities are discussed below in the individual Functions discussion. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.Main Steam Line Isolation l.a.Reactor Vessel Water Level-Low Low Low Level 1 Low reactor pressure vessel (RPV)water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded.The Reactor Vessel Water Level-Low Low Low, Level 1 Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals.The Reactor Vessel Water Level-Low Low Low, Level 1 Function associated with isolation is assumed in the analysis of the recirculation line break (Ref.1).The isolation of the MSLs on Level 1 supports actions to ensure that offsite dose limits are not exceeded for a DBA.Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)(continued),BFN-UNIT 2 B 3.3-147 AMENDMENT ggi il~ Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE SAFETY.ANALYSES LCO, and APPLICABILITY l.a.Reactor Vesse Water Level-Low Low Low Leve 1 (continued) and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-Low Low Low, Level 1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Reactor Vessel'Water Level-Low Low Low, Level 1 Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LCO 3.3.5.1)to ensure that the HSLs isolate on a potential loss of coolant accident (LOCA)to prevent offsite doses from exceeding 10 CFR 100 limits.This Function isolates the Group 1 valves.1.b.Main Steam Line Pressure-Low Low MSL pressure with the reactor at power indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100'F/hr if the pressure loss is allowed to continue.The Main Steam Line Pressure-Low Function is directly assumed in the analysis of the pressure regulator failure (Ref.2).For this event, the closure of the HSIVs ensures that the RPV temperature change.limit (100'F/hr) is not reached.In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded.(This Function closes the HSIVs prior to pressure decreasing below 785 psig, which results in a scram due to HSIV closure, thus reducing reactor power to<25%RTP.)The MSL low pressure signals are initiated from four transmitters that are connected to the HSL header.The transmitters are arranged such that, even.though physically separated from each other, each transmitter is able to detect low MSL pressure.Four channels of Hain Steam Line Pressure-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.(continued) BFN-UNIT 2 B 3.3-148 AMENDMENT ili il~ Primary Containment Isolation Instrumentation 0 B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY I.b.ain Steam ne Pressure-Low (continued) The Main Steam Line Pressure-Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref.2).This Function isolates the Group 1 valves..c.Main Steam'ne F ow-H'ain Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs.If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover.If the RPV water level decreases too far, fuel damage could occur.Therefore, the isolation is initiated on high flow to prevent or minimize core damage.The Main.Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (HSL'B)(Ref.2)..The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not.exceed the 10 CFR 100 limits.The HSL flow signals ar'e initiated from 16 transmitters that are~connected to the four MSLs.The transmitters are arranged such that, even though physically separated from each other, all four connected to one HSL would be able to detect the high flow.Four channels of Hain Steam Line Flow-High Function for each HSL (two channels per trip system)are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break.This Function isolates the Group I valves.1.d.Main Steam Line S ace Tem erature-Hi h The Main Steam Line Space Temperature Function is provided to detect a leak in the RCPB and provides diversity to the high flow instrumentation. The isolation occurs when-a very small leak has occurred.If the small leak is allowed to continue without isolation, offsite dose limits may,be (continued) BFN-VNIT 2 8 3.3-149 AMENDMENT il Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 1.d.Main Steam Line S ace Tem erature-Hi h (continued) reached.However, credit for these.instruments is not taken in any transient or accident analysis in the FSAR, since bounding analyses are performed for large breaks, such as.HSLBs.Hain Steam Line Space temperature signals are initiated from bimetallic temperature switches located in the area being monitored. Sixteen channels of Hain Steam Line Space Temperature-High Function are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The main steam line space temperature detection system Allowable Value is chosen to detect a leak equivalent to between 1%and lN rated steam flow.These Functions isolate the Group 1 valves.Primar Contai ment Isolation 2.a.Reactor Vessel Mater Level-Low Level 3 Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products.The isolation of the primary containment on Level 3 supports actions to ensure that, offsite dose limits of 10 CFR 100 are not exceeded.The Reactor Vessel Mater Level-Low, Level 3 Function associated with isolation is implicitly assumed in the FSAR analysis as these leakage paths are assumed to be isolated post LOCA.Reactor Vessel Water Level-Low, Level 3 signals are initiated from level.transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-L'ow, Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.(continued) BFN-UNIT 2 B 3.3-150 NENDHENT iki i!~ Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 2.a.Reactor Vessel Water Level-Low Level 3 (continued) The Reactor Vessel Water Level-Low, Level 3 Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.This Function, isolates the Group 2, 3, 6, and 8 valves.2.b.Dr well Pressure-Hi h High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR.100 are not exceeded.The Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the FSAR accident analysis as these leakage paths are assumed to be isolated post LOCA.High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell.Four channels of Drywell Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this may be indicati.ve of a LOCA inside primary containment. This Function isolates the Group 2, 6 and 8 valves.Hi h Pressure Coolant In ection and Reactor Core Isolation Coolin S stems Isolation 3.a.4.a.HPCI and RCIC Steam Line Flow-Hi h Steam Line Flow-High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system.If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover.Therefore, the isolations are initiated on high flow to prevent or minimize core damage.The isolation (continued) BFN-UNIT 2 B 3.3-151 AMENDMENT il~~ Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.a.4.a.PCI a d RCIC Steam Li e 1 ow-i (continued) action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.Specific credit for these Functions is not assumed in any FSAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and HSL breaks.However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.The HPCI and RCIC Steam Line Flow-High signals are initiated from transmitters (two for HPCI and two for RCIC)that are connected to the system steam lines.Two channels of both HPCI and RCIC Steam Line Flow-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains the MSLB event as the bounding event.These Functions isolate the Group 4 and 5 valves, as appropriate. 3.b.4.b.HPCI and RCIC Steam Su 1 ine Pressure-Low Low HSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine.These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR.However, they also provide a diver se signal to indicate a possible system break and provide the only signal which will isolate the steam supply lines for certain pipe breaks.These instruments are included in Technical Specifications (TS)because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet Criterion 4 of the NRC Policy Statement (Ref.7).'he HPCI and RCIC Steam Supply Line Pressure-Low signals are initiated from switches (four for HPCI and four for RCIC)that are connected to the system steam line.Four (continued) BFN-UNIT 2 B 3.3-152 NENDHENT

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY.ANALYSES, LCO, and APPLICABILITY 3.b.4.b.HPCI and RCIC Steam Su 1 Line Pressure-Low (continued) channels of both HPCI and RCIC Steam Supply Line Pressure-Low Functions are available. Each Function is considered to have only one trip system since the output from the logic trips a common relay that initiates the isolations. Only thr ee channels of each Function are required to be OPERABLE.The Allowable Values are selected to be high enough to prevent damage to the system's turbine.These Functions isolate the Group 4 and 5 valves, as appropriate. 3.c.4.c.HPCI and RCIC Turbine Exhaust Dia hra m Pressure-~Hi h High turbine exhaust diaphragm pressure indicates" that the pressure may be too high to continue operation of the associated system's turbine.That is, one of two exhaust diaphragms has ruptured and pressure is reaching turbine casing pressure limits..These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR.These instruments are included in the TS because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet Criterion 4 of the NRC Policy Statement (Ref.7).The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High signals are initiated from switches (four for HPCI and four for RCIC)that are connected to the area between the rupture diaphragms on each system's turbine exhaust line.Four channels of both HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available. Each Function is considered to have only one trip system since the output from the logic trips a common relay that initiates the isolations. Only three channels of each Function are required to be OPERABLE.The Allowable Values are low enough to prevent damage to the systems'urbines.(continued) BFN-UNIT 2 B 3.3-153 Amendment ili i Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.c.4.c.PCI and RCIC Turbine Exhaust Dia hr (""'""4)These Functions isolate the Group 4 and 5 valves, as appropriate. 3.d.3.e.4.d.4.e.Area Tem erature-Hi h Area temperatures are provided to detect a leak from the associated system steam piping.The isolation occurs when a very small leak has occur red and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached.These Functions are not assumed in any FSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.Area Temperature-High signals are initiated from bimetallic temperature switches that are appropriately located to protect the system that is being monitored. Four instruments monitor each area.Four channels for each HPCI and RCIC Area and Differential Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Allowable Values are set low enough to detect a leak equivalent to 25 gpm.These Functions isolate the Group 4 and 5 valves, as appropriate. Reactor Water Cleanu S stem Isolation 5.a.S.b.5.c.5.d 5.e.S.f.Area Tem erature-i h RWCU area temperatures are provided to detect a leak from the RWCU System.The isolation occurs even when very small leaks have occurred.If the small leak continues without isolation, offsite dose limits may be reached.Credit for these instruments is not taken in any transient or accident analysis in the FSAR, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.Area temperature signals are.initiated from temperature elements that are located in the room that is being (continued) BFN-UNIT 2 B 3.3-154 AMENDMENT 0 i 0 Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 5.a.5.b.5.c.5.d 5.e.5.f.Area Tem erature-'(continued) monitored. Four sensors in each area are required to be OPERABLE to provide isolation signals to close both RWCU isolation valves using one-out-of-two logic to ensure that no single instrument failure can preclude the isolation function.The Area Temperature-High Allowable Values are set based on the maximum abnormal operating temperature for each area.These Functions isolate the Group 3 valves.5..SLC S stem Initiation The isolation of the RWCU System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the RWCU System (Ref.4).An isolation signal for both RWCU isolation valves is initiated when the SLC pump start handswitch is not in the stop position.There is no Allowable Value associated with this Function since the channels.are mechanically actuated based solely on the position of the SLC System initiation switch.The SLC System Initiation Function is required to be OPERABLE only in NODES 1 and 2, since these are the only NODES where the reactor can be critical, and these NODES are consistent with the Applicability for the SLC System (LCO 3.1.7).As noted (footnote (a)to Table 3.3.6.1-1), the SLC initiation signal provides input to the isolation logic for both RWCU isolation valves.5.h.Reactor Vessel Water Level-Low eve 3 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break.The isolation of the RWCU System on Level 3 supports actions to ensure that the fuel peak cladding temperature remains below the limits: of (continued) BFN-UNIT 2 B 3.3-155 AMENDMENT il~~i il Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 5.h.Reactor Vessel Water Level-Low Level 3 (continued) 10 CFR 50.46.The Reactor Vessel Water Level-Low, Level 3 Function associated with RWCU isolation is not directly assumed in the FSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting). Reactor Vessel Mater Level-Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Mater Level-Low, Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.This Function isolates the Group 2,'3, 6 and 8 valves.Shutdown Coolin S stem Isolation 6.a.Reactor Steam Dome Pressure-Hi h The Reactor Steam Dome Pressure-High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR)System.This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and.credit for the interlock is not assumed in the accident or transient analysis in the FSAR.The Reactor Steam Dome Pressure-High signals are initiated from two switches that are connected to different taps on the RPV.Two channels of Reactor Steam, Dome Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Function is only required to be OPERABLE in NODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed.The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization. This Function isolates Group 2 RHR SDC isolation valves.(continued) BFN-UNIT 2 B 3.3-156 NENDMENT gg~il~ib Primary Containment Isolation Instrumentation B 3.3.6.1 BASES (continued) ACTIONS A Note.has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such,'a Note has been provided that allows separate Condition entry for each inoperable primary containment isolation instrumentation channel.A.l and A.2 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 2.a, 2.b, and 5.h;24 hours for Functions other than Functions 1.d, 2.a, 2.b, and 5.h;and 30 days for Function 1.d has been shown to be acceptable (Refs.5 and 6)to permit restoration of any inoperable channel to OPERABLE status.Required Actions A.1 and A.2 are modified by Notes that specify the Applicability of the Required Actions for Function 1.d when 15 of 16 channels are OPERABLE.Required Action A.2 provides an allowable out of ser vice time of 30 days for Function 1.d when 15 of 16 channels are OPERABLE.This has been shown to be acceptable (Ref.9)to permit restoration of the one inoperable channel to OPERABLE status.This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases).If.the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1 or A.2.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable (continued) BFN-UNIT 2 B 3.3-157 Amendment il~ik Primary Containment Isolation Instrumentation B 3.3.6.1 ACTIONS A.l and A.2 (continued) channel in trip would result in an isolation), Condition C must be entered and its Required Action taken.B.I Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant isolation capability being lost for the associated penetration flow path(s).For MSL, Primary Containment, HPCI, RCIC, RWCU and SOC Isolation Functions where actuation of both trip systems is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip).This ensures that both trip systems will generate a trip signal from the given Function on a valid signal.For those Primary Containment,.HPCI, RCIC, RWCU, and SDC isolation functions, where actuation of one trip system is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in.trip, such that one trip system will generate a trip signal from the given Function on a valid signal.This ensures that at least one of the PCIVs in the associated penetration flow path can receive an isolation signal from the given Function.For all Functions except l.c, l.d, 3.a, 4.a, 5.a through 5.g, and 6.a, this would require both trip systems to have one channel OPERABLE or in trip.For Function 1.c, this would require both trip systems to have one channel, associated with each NSL, OPERABLE or in trip.For Function 1.d, which consists of channels that monitor several locations within a given area (e.g., different locations within the main steam tunnel area), this would require both trip systems to have one channel per location OPERABLE or in trip.For Functions 3.a, 4.a, and 6.a, this would require one trip system to have one channel OPERABLE or in trip.For Functions 5.a through 5.f, this would require both trip systems to have one channel, associated with each area, OPERABLE or in trip.The Completion Time is, intended to allow the operator time to evaluate and repair any.discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes (continued) BFN-UNIT 2 B 3.3-158 Amendment ili il~0 Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTION B.l (continued) risk while allowing time for restoration or tripping of channels.The second Completion Time for Function 1.d when normal ventilation is not available is provided to allow the plant to avoid an HSL isolation transient when recovering from a temporary loss of ventilation in the HSL tunnel area (e.g., during performance of the secondary containment leak rate tests).As allowed by LCO 3.0.2 (and discussed in the Bases for LCO 3.0.2), the plant may intentionally enter this condition to avoid an HSL isolation transient and bypass the high temperature channels during restoration of ventilation flow.However, during the period that multiple Hain Steam Tunnel Temperature -High Function channels are inoperable due to this intentional action, an additional compensatory measure is deemed necessary and shall be taken: an operator shall observe control room indications of the affected space temperatures for indications of'mall steam leaks.In the event of rapid increases in temperature (indicative of a steam line break), the operator shall promptly close the HSIVs.The 4 hour Completion Time is acceptable because along with the compensatory measures described above it minimizes risk while allowing time for restoration or tripping of channels.C.1 Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1.The applicable Condition specified in Table 3.3.6.1-1 is Function and HODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A or B and the associated Completion Time has expired, Condition C will be entered for that channel and.provides for transfer to the appropriate subsequent Condition. D.1 D.2.1 and D.2.2 If;the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must (continued). BFN-UNIT 2 B 3.3-159 Amendment ili Oi 0 Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS.1 D.2.1 and D.2.2 (continued) be placed in a MODE or other specified condition in which the LCO does not apply.This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and D.2.2).Alternately, the associated MSLs may be isolated (Required Action D.l), and, if allowed (i.e., plant safety analysis allows operation wi.th an MSL isolated), operation with that MSL isolated may continue.Isolating the affected MSL accomplishes the safety function of the inoperable channel.The 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.E.1 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.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, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.F.l If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path(s)is isolated.Isolating the affected penetration flow path(s)accomplishes the safety function of the inoperable channels.For the RWCU Area Temperature-High Functions, the affected penetration flow path(s)may be considered isolated by isolating only that portion of the system in the associated room monitored by the inoperable channel.That is, if the RWCU pump room A area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing the B RWCU pump.(continued) BFN-UNIT 2 B 3.3-160 AMENDMENT il~ili il Primary Containment Isolation Instrumentation 8 3.3.6.1 ACTIONS F.1 (continued) Alternately, if it is not desired to isolate the affected penetration flow path(s)(e.g., as in the case where isolating the penetration flow path(s)could result in a reactor scram), Condition G must be entered and its'Required Actions taken.The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path(s).G.l and G.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or any Required Action of Condition F is not met and the associated Completion Time has expired, the plant must be placed in, a MODE or other specified condition in which the LCO does not apply.This is done by placing the plant in at least MODE 3 within 12 hours and in'MODE 4 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.H.1 and H.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the SLC System is declared inoperable or the RWCU System is isolated.Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the SLC System inoperable or isolating the RWCU System.The 1 hour Completion Time is" acceptable because it minimizes risk while allowing sufficient time for personnel to,isolate the RWCU System.(continued) BFN-UNIT 2 B 3.3-161 AMENDMENT ili$Q>>0 Primary Containment Isolation Instrumentation B 3.3.6.1 BASES (continued) SURVEILLANCE REQUIREMENTS As noted (Note 1)at the beginning of the-SRs, the SRs for each Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1. The Surveillances are modified by a Note (Note 2)to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions: and Required Actions may be delayed for up to 6 hours provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This Note is based on the reliabil.ity analysis (Refs.5 and 6)assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetr ation flow path(s)when necessary. SR 3.3.6.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. J Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency is based on operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of (continued) BFN-UNIT 2 B 3.3-162 NENDNENT ~i ili Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.6.l.1 (continued) channels during normal operational use of the displays associated with the channels required by the LCO.SR 3.3.6.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.6.1.2 is based on the reliability analysis described.in References 5 and 6.0 SR 3.3.6.1.3 SR 3.3.6.1.4 and SR 3.3.6.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequencies of SR 3.3.6.1.3, SR 3.3.6.1.4, and SR 3.3.6.1.5 are based on the magnitude of equipment drift in the setpoint analysis.SR 3.3.6.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function.The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed'with the reactor at power.(continued) BFN-UNIT 2 B 3.3-163 AMENDMENT ili ggi Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE REQUIREMENTS Operating experience has shown these components usually pass the Surveillance when performed at the Frequency provided.REFERENCES 1.FSAR, Section 6.5.2.FSAR, Chapter 14.3.NED0-31466,"Technical Specification Screening Criteria Application and Risk Assessment,'" November 1987.4., FSAR, Section 4.9.3.5.NEDC-31677P-A,"Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.6.NEDC-30851P-A Supplement 2,"Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.7.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.8.FSAR, Section 5.2.9.NRC letter from Richard J.Clark to Hugh G.Parris dated August 9, 1984, Safety Evaluation for Amendment Nos.107, 101, and 74 to Facility Operating License Nos.DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant Units 1, 2, and 3 respectively. BFN-UNIT 2 B 3.3-164 AMENDMENT <gi il~ Secondary Containment Isolation Instrumentation B 3.3.6.2 B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs)and starts the Standby Gas Treatment (SGT)System.The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs)(Ref.I).Secondary containment isolation and establishment of vacuum with the SGT System within the assumed time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits.The isolation. instrumentation includes the sensors, relays, and.switches that are necessary to cause initiation of secondary containment isolation. Most channels include electronic equipment (e.g., trip units)that compares measured input signals with pre-established setpoints. Mhen the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic.Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (I)reactor vessel water level, (2)drywell pressure, (3)reactor zone exhaust high radiation, and (4)refueling floor exhaust high radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. In addition, manual initiation of the logic is provided.The output signals from the secondary containment isolation logic isolates secondary containment and starts all three SGT subsystems to provide for the necessary filtration of fission products.BFN-UNIT 2 B 3.3-165 (continued) AMENDMENT lyly O~Ik Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ,(continued) APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The isolation signals generated by the secondary containment isolation instrumentation are implicitly assumed in the safety analyses of References 1 and 2 to initiate closure of valves and start the SGT System to limit offsite doses.Refer to LCO 3.6.4.2,"Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3,"Standby Gas Treatment (SGT)System," Applicable Safety Analyses Bases for more detail of the safety analyses.The secondary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.7).Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set.within the specified Allowable Values, as shown in Table 3.3.6.2-1. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.Each..channel must also respond.within its assumed response time, where appropriate. Allowable Values are specified for each Function specified in the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.(continued) BFN-UNIT 2 B 3.3-166 AMENDMENT hli i Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO,, and APPLICABILITY (continued) The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances,, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.In general, the individual Functions are required to be OPERABLE in the NODES or other specified conditions when SCIVs and the SGT System are required.The speci.fic Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.1.Reactor Vessel Water Level-Low Level 3 Low reactor pressure vessel (RPV)water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release.The Reactor Vessel Water Level-Low, Level 3 Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals.The isolation and initiation systems on Reactor Vessel Water Level-Low, Level 3 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Ref.4).Reactor Vessel Water Level-Low, Level 3 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-Low, Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.(continued) BFN-UNIT 2 8 3.3-167 NENDMENT ik~~i Secondary Containment Isolation Instrumentation B 3.3.6.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 1.Reactor Vessel Water Level-Low Low Level (continued) The Reactor Vessel Water Level-Low, Level 3 Function is required to be OPERABLE in MODES I, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS);thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas.In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES;thus, this Function is not required.In addition, the Function is also required to be OPERABLE during.operations with a potential for draining the reactor vessel (OPDRVs)because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs.2.Dr well Pressure-Hi h High drywell pressure can indicate a break in.the reactor coolant pressure boundary (RCPB).'n isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release.The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.However, the Drywell Pressure-High Function associated with isolation is not assumed in any FSAR accident or transient analyses.It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell.Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (continued) BFN-UNIT 2 B 3.3-168 AMENDMENT il 0 0 Secondary Containment Isolation Instrumentation 8 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 2.Dr ell Pressure-Hi h (continued)(LCO 3.3.5.1)since this is indicative of a loss of coolant accident (LOCA).The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS;thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas.This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.3 4.Reactor Zone and Refuelin Floor Exhaust Radiation-Hi h High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident.When Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the FSAR safety analyses (Ref.4).The Exhaust Radiation-High signals are initiated from radiation detectors located on the ventilation exhausts coming from each reactor zone and the common refueling zone.There are two radiation monitors for each ventilation exhaust path.There are two pairs of radiation elements which monitor the ventilation exhaust from each zone.Each pair of radiation elements provides input to one radiation monitor.Both radiation elements must provide a High signal to trip the.associated radiation monitor (two-out-of-two). However, if either radiation monitor trips, a secondary containment isolation signal is initiated (one-out-of-two). Two channels (monitors) of Reactor Zone Exhaust Radiation-High Function and two channels of Refueling Floor Exhaust Radiation-High Function are available and are required'o be OPERABLE to ensure that no single instrument failure can preclude the isolation function.There is only one trip system for each Function.(continued) BFN-UNIT 2 B 3.3-169 AMENDMENT iS~il 4I Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3 4.Reactor Zone and efuel n Floor x aust Radiation-Hi h (continued) The Allowable Values are chosen to provide timely detection of nuclear system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolation. The Reactor Zone and Refueling Floor Exhaust Radiation-High Functions are required to be OPERABLE in MODES I, 2, and 3 where considerable energy exists;thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas.In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES;thus, these Functions are not required.In addition, the Functions are also required to be OPERABLE during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded.ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.(continued) BFN-UNIT 2 8 3.3-170 AMENDMENT 4l Secondary Containment Isolation Instrumentation B 3.3.6.2 ACTIONS (continued) Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours fo}Functions 1 and 2, and 24 hours for Functions other than Functions 1 and 2, has been shown to be acceptable (Refs.5 and 6)to permit restoration of any inoperable channel to OPERABLE status.This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.l Bases).If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.l.Placing the.inoperable channel in trip would conservatively compensate for the inoperability, restore.capability to accommodate a single failure, and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in.the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Actions taken.B.1 Required Action B.l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated secondary containment penetration flow path(s)or a complete loss of automatic initiation capability.for the SGT System.A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal.This ensures that one of the two SCIVs in the associated penetration flow path and two SGT subsystems can be initiated on an isolation signal from the given Function.For Functions with two one-out-of-two logic trip systems (Functions 1 and 2), this would require one trip system to have one channel OPERABLE or in trip.For Functions with one one-out-of-two logic trip system (Functions 3 and 4), this would require the trip system to have one channel.OPERABLE or in trip.(continued).BFN-UNIT 2 B 3.3-171 AMENDMENT ll!~0 Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS B.1 (continued) The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.C.l.l C..2 C.2.1 and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured.Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function.Isolating the associated zone (closing the ventilation supply and exhaust automatic isolation dampers)and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1)performs the intended function of the instrumentation and allows operation to continue.Alternately, declaring the associated SCIVs or SGT subsystem(s) inoperable (Required Actions C.l.2.and C.2.2)is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3)provide appropriate actions for the inoperable components. One hour is sufficient for plant operations. personnel to establish required plant conditions or to declare the assoc'iated components inoperable without unnecessarily challenging plant systems.SURVEILLANCE RE(UIREMENTS As noted (Note 1)at the beginning of the SRs, the SRs for each Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1. The Surveillances are modified by a Note (Note 2)to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour (continued) BFN-UNIT 2 B 3.3-172 NENDHENT 4I~'IL II Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) allowance, the channel must be returned to OPERABLE status'r the applicable Condition entered and Required Actions taken.This Note is based on the:reliability analysis (Refs.5 and 6)assumption of the average time required to perform channel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary. The Surveillances are modified by a third Note (Note 3)to indicate that for Functions 2.c and 2.d, when a channel is placed in an inoperable status solely for performance of required testing or maintenance,, entry into associated Conditions and Required Actions may be delayed for up to 6 hours for a CHANNEL FUNCTIONAL TEST and for up to 24 hours for a CHANNEL CALIBRATION or maintenance, provided the downscale trip of the inoperable channel is placed in the tripped condition. Upon completion of the Surveillance or maintenance, or expiration of the 6 hour or 24 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between, each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.(continued) BFN-UNIT 2 B 3.3-173 NENDMENT 0 0 0 Secondary Containment Isolation Instrumentation B 3.3.6.2 SURVEILLANCE REQUIREMENTS The Frequency is based on operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.S 3.3.6.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6.SR 3.3.6.2.3 and SR 3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.The system functional testing performed on SCIVs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.The 18 month Frequency for Functions 1 and 2 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.The 184 day Frequency for Functions 3 and 4 is based on operating.experience and equipment capability. Operating experience has shown that these components usually pass the Surveillance when performed at these Frequencies. Therefore, the Frequencies were found to be acceptable from a reliability standpoint.(continued) BFN-UNIT 2 B 3.3-174 AMENDMENT il ii Secondary Containment Isolation Instrumentation B 3.3.6.2-BASES SURVEILLANCE 'REQUIREMENTS A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency of SR 3.3.6.2.4.is based on the magnitude.of equipment drift in the setpoint analysis.REFERENCES 1.FSAR, Chapter 5 and Section 7.3.5-.2.FSAR,, Chapter 14.3.FSAR, Section 14.6.3.5.4.FSAR, Sections 14.6.3.6 and 14.6.4.5.5.NEDC-31677P-A,"Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.6.NEDC-30851P-A Supplement 2,"Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and,ECCS Instrumentation," March 1989.7.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-175 AMENDMENT il 0 Cl CREV System Instrumentation B 3.3.7.1 B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Ventilation (CREV)System Instrumentation BASES BACKGROUND The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function.The instrumentation and controls for the CREV System automatically initiate action to pressurize the control room (CR)to minimize the consequences of radioactive material in the control room environment. In the event of a Reactor Vessel Water Level-Low, Level 3, Drywell Pressure-High, Reactor Zone Exhaust Radiation-High, Refueling Floor Exhaust Radiation-High, or Control Room Air Supply Duct Radiation-High signal, the CREV System is automatically started, in the pressurization mode.The air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the CR slightly pressurized. The CREV System instrumentation has one or two trip systems, which can initiate both CREV subsystems (only the selected subsystem will be initiated)(Ref.1).Each trip system receives input from each of the Functions listed above.The Functions are arranged as follows for each trip system.The Reactor Vessel Water Level-Low, Level 3 and Drywell Pressure-High are each arranged in a one-out-of-two taken twice logic (these signals are the same that isolate the primary containment). The Reactor Zone Exhaust Radiation-High,.Refueling Floor Exhaust Radiation-High and Control Room Air Supply Duct Radiation-High (only one trip system)are each arranged in a one-out-of-two logic.The channels include electroni.c equipment (e.g., trip relays)that compares me'asured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation.signal to the initiation logic.BFN-UNIT 2 B 3.3-176 (continued) Amendment ll'Cl CREV System Instrumentation B 3.3.7.1 BASES (continued) APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The ability of the CREV System to maintain the habitability of the CR is explicitly assumed for certain accidents as discussed in the FSAR safety analyses (Ref.,2).CREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits s'et by GDC 19 of 10 CFR 50, Appendix A.CREV System instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.5).The OPERABILITY of the CREV System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.7.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified, Allowable Values, where appropriate. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.The setpoint is calibrated consistent with applicable setpoint methodology assumptions 'nominal trip.setpoint). Allowable Values are specified for each CREV System Function specified in the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from-the safety analysis.The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for (continued) BFN-UNIT 2 B 3.3-177 NENDNENT 0 0 4b CREV System Instrumentation B 3.3.7.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for., The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.Reactor Vesse Water Level-Low Level 3 Low reactor pressure vessel (RPV)water level indicates that the capability of cooling the fuel may be threatened. A low reactor vessel water level could indicate a LOCA and will automatically initiate the CREV System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel. Reactor Vessel Water Level-Low, Level 3.signals are initiated from four level transmitters that sense the difference between the, pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg)in the vessel.Four channels of Reactor Vessel Water Level-Low, Level 3 Function are available (two channels per trip system).and are required to be OPERABLE to ensure that a single instrument failure cannot preclude CREV System initiation. The Reactor Vessel Water Level-Low, Level 3 instrumentation which provides input signals to the CREV System initiation logic is the same instrumentation which provides the input signals for the Primary Containment Isolation System logic (LCO 3.3.6.1).The Reactor Vessel Water Level-Low, Level 3 Function is required to be OPERABLE in MODES 1, 2, and 3, and during operations with a potential for draining the reactor vessel (OPDRVs)to ensure that the control room personnel are protected during a LOCA.In MODES 4 and 5 at times other than OPDRVs, the probability of a vessel draindown event resulting in a release of radioactive material into the environment is minimal.In addition, adequate protection is performed by the Control, Room Air Supply Duct Radiation-High Function.Therefore, this Function is not required in other MODES and specified conditions.(continued) BFN-UNIT 2 B 3.3-178 AMENDMENT il~ib Cl CREV System Instrumentation B 3.3.7.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.Dr wel Pressure-Hi h High pressure in the drywell could indicate a break in the reactor coolant pressure boundary.A high drywell pressure signal could indicate a LOCA and will automatically initiate the CREV System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel. Drywell Pressure-High signals are initiated from four pressure transmitters that sense drywell pressure.Four channels of Drywell Pressure-High Function are available (two channels per trip system)and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System initiation. The Drywell Pressure-High Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Allowable Value (LCO 3.3.5.1).The Drywell Pressure-High Function is required to be OPERABLE in MODES I, 2, and 3 to ensure that control room personnel are protected in the event of a,LOCA.In NODES 4 and 5, the Drywell Pressure-High Function is not required since there is insufficient energy in the reactor to pressurize the drywell to the Drywell Pressure-High setpoint.3.4.Reactor Zone and efuelin Flop Exhaust Radiatio-Hi h High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.The release may have originated from the, primary containment due to a break in the RCPB.when Exhaust Radiation-High is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fission products.Additionally, high radiation in the refueling floor exhaust could be the result of a fuel handling accident.A reactor zone or refueling floor exhaust high radiation signal will automatically initiate the CREV System, since this radiation release could result in radiation exposure to control room personnel. The reactor zone and refueling floor exhaust radiation equipment consists of two independent monitors and channels (continued) BFN-UNIT 2 B 3.3-179 NENDMENT il 0 CREV System Instrumentation B 3.3.7.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 3.4.eacto o e and Refuelin Flop.Exh ust located on the ventilation exhaust piping coming from the reactor building and the refueling zones, respectively. Two channels of each function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System initiation. There is only one trip system for each Function.The Allowable Value was selected to ensure that the Function will promptly detect high activity that could threaten exposure to control room personnel. The Reactor Zone and Refueling Floor Exhaust Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 and during movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and operations with a potential for draining the reactor vessel (OPDRVs), to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event.During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA or fuel damage is low;thus, the Function is not required.5.Control Room Air Su l Duct Radiation-Hi h The control room air supply duct radiation monitors measure radiation levels exterior to the inlet ducting of the CR.A high radiation level may pose a threat to CR personnel; thus, the CREV System is automatically initiated on a control room air supply duct high radiation signal.The Control Room Air Supply Duct Radiation-High Function consists of two independent monitors.Two channels of Control Room Air Supply Duct Radiation-High are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System.initiation. There is only one trip system for this Function.The Allowable Value was selected to ensure protection of the control room personnel. The Control Room Air Supply Duct Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel (continued) BFN-UNIT 2 B 3.3-180 AMENDMENT

CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 5.Control Room Air Su l Duct Radiation-Hi h (continued) assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handl-ing event, or vessel draindown event.During NODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA or fuel damage is low;thus, the Function is not required.ACTIONS A Note has been provided to modify the ACTIONS related to CREV System instrumentation, channels.Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, 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. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel.A.1 Required Action A.l directs entry into the appropriate Condition referenced in Table 3.3.7.1-1. The applicable Condition specified in the Table is Function dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. B.and B.2 Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 12 hours has been shown to be acceptable (Refs.3 and 4)to permit restoration of any inoperable channel to OPERABLE status.(continued) BFN-UNIT 2 B 3.3-.181 NENDNENT 0, 0 CREV System Instrumentation B 3.3.7.1 ACTIONS B.l and B.2 (continued) However, this out of service time is only acceptable provided the associated Function is still maintaining CREV System initiation capability. A Function is considered to be maintaining-CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal from the given Function on a valid signal.In this situation (loss of CREV System initiation capability), the 12 hour allowance of Required Action 8.2 is not appropriate. If the Function is not maintaining CREV System initiation capability, the CREV System must be declared inoperable within 1 hour of discovery of the loss of CREV System initiation capability in both trip systems.The 1 hour Completion Time (B.1)is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.If.the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.C.l and C.2 Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 24 hours is provided to permit restoration of any inoperable channel to'OPERABLE status.However, this out of service time is only-acceptable provided the associated Function.is still maintaining CREV System initiation capability. In this situation (loss of CREV System initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate. If the Function is not maintaining CREV System initiation capability, the CREV System must be declared inoperable (continued) BFN-UNIT 2 B 3.3-182 NENOMENT il~0 CREV System Instrumentation B 3.3.7.1 BASES ACTIONS C I and C.2 (continued) within 1 hour of discovery of the loss of CREV System initiation capability in both trip systems.The 1 hour Completion Time (C.l)is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action C.2.Placing the inoperable channel in trip performs the intended function of the channel (starts the selected CREV subsystem in the pressurization mode).Alternately, if it is not desired to place the channel in trip (e.g., as in the case where it is not desired to start the subsystem), Condition E must be entered and its Required Action taken.D.l D.2 and D.3 Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREV System design, Required Action D.1 allows continued operation with an inoperable channel provided repair is initiated in a timely manner and the remaining OPERABLE channel is functionally tested once per 24 hours.With two channels of the Control Room Air Supply Duct Radiation-High function inoperable (Required Actions D.2 and D.3), an allowed outage time of 30 days is provided to restore at least one channel to OPERABLE status provided that the alternate monitoring capability is verified functional once per 12 hours.The alternate monitoring capability is provided by the control room particulate monitor (RN-90-53) and radiation monitor (RE-90-8). These monitors alarm in the control.room on high activity.Upon receipt of these alarms, the operator is required to manually isolate the control room and manually initiate the emergency pressurization system.The 30 day allowed outage time is based on verifying functional capability of these two monitors and the administrative controls that require operator action to manually initiate a CREV subsystem.(continued) BFN-UNIT 2 B 3.3-183 AMENDMENT !l~il CREV System Instrumentation B 3.3.7.1 BASES ACTIONS (continued) E.l and E.2 Mith any Required Action and associated Completion Time not met, the associated CREV subsystem(s) must be placed in the pressurization mode of operation per Required Action E.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident.The method used to place the CREV subsystem(s) in operation must provide for automatically re-initiating the subsystem(s) upon restoration of power following a loss of power to the CREV subsystem(s). Alternately, if it is not desired to start the subsystem(s), the CREV subsystem(s) associated with inoperable, untripped channels must be declared inoperable within 1 hour.The 1=hour Completion Time is intended to allow the operator time to place the CREV subsystem(s) in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, for placing the associated CREV subsystem(s) in operation, or.for entering the applicable Conditions and Required Actions for the inoperable CREV subsystem(s). SURVEILLANCE REQUIREMENTS As noted (Note 1)at the beginning of the SRs, the SRs for each CREV System instrumentation Function are located in the SRs column of Table 3.3.7.1-1. The Surveillances are modified by a Note (Note 2)to indicate that when a channel-is placed in an inoperable status solely for performance of required Sur veillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.This, Note is based on the reliability analysis (Refs.3 and 4)assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREV System will initiate when necessary.(continued) BFN-UNIT 2 B 3.3-184 AMENDMENT il Cl CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE REQUIREMENTS (continued) The Surveillances are modified by a third Note (Note 3)to indicate that for Functions 3 and 4, when a channel is placed in an inoperable status solely for performance of required testing or maintenance, entry into associated Conditions and Required Actions may be delayed for up to 6 hours for a CHANNEL FUNCTIONAL TEST and for up to 24 hours for a CHANNEL CALIBRATION or maintenance, provided the downscale trip of the inoperable channel is placed in the tripped condition. Upon completion of the Surveillance or maintenance, or expiration of the 6 hour or 24 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.SR 3.3.7.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred.A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.It is based on the assumption that instrument channels monitoring the same parameter should, read approximately the same value.Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.A CHANNEL CHECK will detect gross channel failure;thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.The Frequency is based upon operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.(continued) BFN-UNIT 2 B 3.3-185 NENDNENT il~ CREV System Instrumentation B 3.3.7.1 SURVEILLANCE REQUIREMENTS (continued) SR 3.3.7.1.2'A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analyses of References 3 and 4.SR 3.3.7.1.3 and SR 3.3.7.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequencies are based upon the magnitude of equipment drift in the setpoint analysis.SR 3.3.7.1.4 and SR 3.3.7.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel.The system.functional testing performed in LCO 3.7.3,"Control Room Emergency Ventilation (CREV)System," overlaps this Surveillance to provide complete testing of the assumed safety function.The 184 day Frequency for Function 5 is based on equipment capability. The 18 month Frequency for Functions 1, 2, 3, and 4 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has shown these components usually pass the Surveillance when performed at their designated Frequencies. BFN-UNIT 2 B 3.3-186 (continued) AMENDMENT il CREV System Instrumentation B 3.3.7.1 BASES (continued) REFERENCES 1.FSAR;Section 10.12.5.3. 2.FSAR,, Section 14.6.3.7.3.GENE-770-06-1,"Bases for Changes to Survei.llance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.4.NEDC-31677P-A,"Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.5.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.0 BFN-UNIT'2 B 3.3-187 NENDMENT 0 l 0 Cl LOP Instrumentation B 3.3.8.1 B 3.'3 INSTRUMENTATION B 3.3.8.1 Loss of Power (LOP)Instrumentation BASES BACKGROUND Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS)is dependent upon the availability of'dequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV shutdown boards.Offsite power is the preferred source of power for the 4;16 kV shutdown boards.If the monitors determine that insufficient power is available, the boards are disconnected from the offsite power sources and connected to the onsite diesel generator (DG)power sources.Each 4.16 kV shutdown board has its own independent LOP instrumentation and associated trip logic.The voltage for each board is monitored at two levels, which can be considered as.two different undervoltage Functions: Loss of Voltage and 4.16 kV Shutdown Board Undervoltage Degraded Voltage.Each Function causes various board transfers and disconnects. C The Degraded Voltage Function is monitored by three undervoltage relays for each shutdown board, whose outputs are arranged in a two-out-of-three logic configuration (Ref.1).The channels include electronic equipment (e.g., trip relays)that compare measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay deenergizes, which then outputs a LOP trip signal to the shutdown board logic.The Loss of Voltage Function is monitored by two undervoltage relay pairs for each shutdown board, where outputs are arranged in a two-out-of-two logic configuration (Ref.1-).The channels include four electro-mechanical relays, two of which must deenergize to start the associated diesel generator and another two which must deenergize to initiate load shed of the associated 4.16 kV shutdown board.BFN-UNIT 2 B 3.3-188 (continued) AMENDMENT i LOP Instrumentation B 3.3.8.1 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The LOP instrumentation is required for Engineered Safety Features to function in any accident with a loss of offsite power.The required channels of LOP instrumentation ensure that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 2, 3, and 4 analyzed accidents in which a loss of offsite power is assumed.The initiation of the DGs on loss of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.Accident analyses credit the loading of the DG based on the loss of offsite power concurrent with a loss of coolant accident.The diesel starting and loading times have been included.in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.The LOP instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref.5).The OPERABILITY of the LOP instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.8.1-1.Each Function must have a required number of OPERABLE channels per 4.16 kV shutdown board, with their setpoints within the specified Allowable Values.A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. The Allowable Values are specified for each Function in the Table.Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within the Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., degraded voltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay)changes state.The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.The Allowable Values are derived from the (continued) BFN-UNIT 2 B 3.3-189 NENDMENT il~il 41 LOP Instrumentation B 3.3.8.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) analytic limits, corrected for calibration, process, and some of the instrument errors.The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift).The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for unit channels that must function in harsh environments as defined by 10 CFR 50.49)are accounted for.The specific Applicable Safety Analyses, LCO, and.Applicability discussions are listed below on a Function by Function basis.4.16 kV Shutdown Board Undervolta e Loss of Volta e Loss of voltage on a 4.16 kV shutdown board indicates that:offsite power may be completely lost to the respective shutdown board and is unable to supply sufficient power for proper operation of the applicable equipment. Therefore, the power supply to the board is transferred from offsite power to DG'ower upon total loss of shutdown board voltage for 1.5 seconds.The transfer will not occur if the voltage recovers to the specified Allowable Value for Reset Voltage within 1.5 seconds.This ensures that adequate power will be available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available to the required equipment. One channel of 4.16 kV Shutdown Board Undervoltage (Loss of Voltage)Function per associated shutdown.board is only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude the DG function.Refer to LCO 3.8.1,"AC Sources-Operating," and 3.8.2,"AC Sources-Shutdown," for Applicability Bases for the DGs.2.4.16 kV Shutdown Board Undervolta e De raded Volta e A reduced voltage condition on a 4.16 kV shutdown board indicates that, while offsite power may not be completely (continued) BFN-UNIT 2 B 3.3-190 NENDMENT il'l' LOP Instrumentation B 3.3.8.1 APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY 2.4.16 kV Shutdown Board Undervolta e De raded Volta e (continued) lost to the respective shutdown board, available power maybe insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.Therefore, power supply to the board is transferred from offsite power to onsite DG power when the voltage on the board drops below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay).This ensures that adequate power will be available to the required equipment. The Board Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient power is available to the required equipment. One channel of 4.16 kV Shutdown Board Undervoltage (Degraded Voltage)Function per associated board is only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude the DG function.Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.ACTIONS A Note has been provided to modify the ACTIONS related to LOP instrumentation channels.Section 1.3, Completion Times, specifies that once a Condition has'been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, wi.ll not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels.As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.(continued) BFN-UNIT 2 B 3.3-191 NENDHENT 0 LOP Instrumentation B 3.3.8.1 BASES ACTIONS (continued) A.l With one of the three phase-to-phase degraded voltage relays inoperable, Required Action A.1 provides a 15 day allowable out of service time to restore the relay to OPERABLE status.The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays.If the inoperable relay cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.l.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP'nstrumentation), and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition E must be entered and its Required Action taken.B.1 With one or more loss of voltage relay channels inoperable, the Function is not capable of performing the intended function.Required Action B.1 provides a 10 day allowable out.of service time since the degraded voltage relay channel on the same shutdown board is independent of the loss of voltage relay channel and will continue to function and start the diesel generators on a complete loss of voltage.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.l.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g , as in the case where placing the channel in trip would result in a DG initiation), Condition E must be entered and its Required Action taken.(continued) BFN-UNIT 2 B 3.3-192 Amendment il i~0 LOP Instrumentation B 3.3.8.1 BASES ACTIONS (continued) C.l With one or more degraded voltage relay channels inoperable, the Function is not capable of performing the intended function.Required Action C.1 provides a 10 day allowable out of service time, since the loss of voltage relay channel on the same shutdown board is independent of the degraded voltage relay channel and will continue to function and start the diesel generators on a complete loss of voltage.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required'ction C.l.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition E must be entered and its Required Action taken.D.l and'.2 With the degraded voltage relay channel and the loss of voltage relay channel inoperable on the same shutdown board, the associated diesel generator will not automatically start upon degraded voltage or complete loss of voltage on that shutdown board.In this situation, Required Action D.2 provides a 5 day allowable out of service time-provided the other shutdown boards and undervoltage relays are OPERABLE.Immediate verification of the OPERABILITY of the other shutdown boards and undervoltage relays is therefore required (Required Action D.1).This may be performed as an administrative check by examining logs or other information to determine i'f this equipment is out of service for maintenance or other reasons.It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition E must be entered immediately. The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV Shutdown Boards.The 4.16 kV Shutdown Boards have a similar allowable out of service time.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per (continued) BFN-UNIT 2 B 3.3-193 Amendment Ik!I LOP Instrumentation B 3.3.8.1 ACTIONS Required Action D.2.Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue.Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition E must be entered and its Required Action taken.If any Required Action and associated Completion Time are not met, the associated Function is not capable of performing the intended function.Therefore, the associated DG(s)is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions for the'inoperable DG(s).SURVEILLANCE REQUIREMENTS As noted (Note 1)at the beginning of the SRs, the SRs for each LOP instrumentation Function are located in the SRs column of Table 3.3.8.1-1.The Surveillances are modified by a Note (Note 2)to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided the associated Function maintains initiation capability for three DGs.The loss of function for one DG for this short period is appropriate since only three of four DGs are required to start within the required times and because there is not appreciable impact on risk.Upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.(continued) BFN-UNIT 2 B 3.3-194 AMENDMENT 0 LOP Instrumentation B 3.3.8.1'BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.8.1.1'nd SR 3.3.8.1.2 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant.specific setpoint methodology. Any setpoint adjustment shall be consistent with the.assumptions of the current plant specific setpoint methodology. The Frequency is based upon the calibration interval assumed in the determination of the magnitude of equipment drift in the setpoint analysis.Jl The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY, of the required actuation logic for a specific channel..The system functional testing performed in LCO 3.8.1 and'LCO 3.8.2 overlaps this.Surveillance to provide complete testing of the assumed safety functions. The 18 month Frequency is based on the need to, perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.Operating experience has, shown these components usually pass the Surveil,lance.when performed at the 18 month Frequency. REFERENCES 1.FSAR, Figure 8.4-4.2.FSAR, Section 6.5.3.FSAR, Section 8.5.4.4.FSAR, Chapter-14.5.NRC No.93-102,"Final Policy Statement on Technical Speci.fication Improvements," July 23, 1993.BFN-UNIT 2 B'.3-195 AMENDMENT O~il II RPS Electric Power Monitoring B 3.3.8.2 B 3.3 INSTRUMENTATION B 3.3.8.2 Reactor Protection System (RPS)Electric Power Monitoring BASES BACKGROUND RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motor generator (MG)set or an alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This system protects the loads connected to the'RPS bus against unacceptable voltage and frequency conditions (Ref.1)and forms an important part of the primary success path of the essential safety circuits.Some of the essential equipment powered from the RPS buses includes the RPS logic and scram solenoids. RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the two MG sets or the alternate power supply and will de-energize its respective RPS bus, thereby causing all safety functions normally powered by this bus to de-energize. In the event of failure of an RPS Electric Power Monitoring System (e.g., both in series electric power monitoring assemblies), the RPS loads may experience significant effects from the unmonitored power supply.Deviation from the nominal conditions can potentially cause damage to the scram solenoids and other Class lE devices.In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram action.In the event of an overvoltage condition, the RPS logic relays and scram solenoids may experience a voltage higher than their design voltage.If the overvoltage condition persists for an extended time period, it may cause equipment degradation and the loss of pl'ant safety function.Two redundant Class lE contactors are connected in series between each RPS bus and its MG set, and between each RPS bus and its alternate power supply.Each of these contactor s has an associated independent set of Class lE overvoltage, undervoltage, and underfrequency sensing logic.Together, a contactor and its sensing logic constitute an (continued) BFN-UNIT 2 B 3.3-196 AMENDMENT 0 Cl RPS Electric Power Monitoring B 3.3.8.2 BASES BACKGROUND (continued) electric power monitoring assembly.If the output of the MG set exceeds predetermined limits of overvoltage, undervoltage, or underfrequency, for>4 seconds, a trip relay driven by this 1'ogic circuitry opens the contactor, which removes the associated power supply from service.The timer is common to the three trip relays.APPLICABLE SAFETY ANALYSES The RPS electric power monitoring is necessary to meet the assumptions of the safety analyses by ensuring that the equipment powered from the RPS buses can perform its intended function.RPS electric power monitoring provides protection to the RPS and other systems that receive power from the RPS buses, by acting to disconnect the'RPS from the power supply under specified conditions that could damage the RPS bus powered equipment. RPS electric power monitoring satisfies Criterion 3 of the NRC-Policy Statement (Ref.3).LCO The OPERABILITY of each RPS electric power monitoring assembly is dependent on the OPERABILITY of the overvoltage, undervoltage, and underfrequency logic, as well as the OPERABILITY of the associated contactor. Two electric power monitoring assemblies are required to be OPERABLE for each inservice power supply.This provides redundant protection against any abnormal voltage or frequency conditions to ensure that no single RPS electric power monitoring assembly failure can preclude the function of RPS bus powered components. Each inservice electric power monitoring assembly's trip logic setpoints are required to be within the specified Allowable Value.The actual setpoint is calibrated consistent with applicable setpoint procedures (nominal trip setpoint). Allowable Values are specified for each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected based on engineering judgment and operational experience to ensure that the,setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its (continued) BFN-UNIT 2 B 3.3-197 AMENDMENT l5 ib 41 RPS Electric Power Monitoring B 3.3.8.2 BASES LCO (continued) Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.Trip setpoints are those predetermined values of output at which an action should take place.The setpoints are compared to the actual process parameter (e.g., overvoltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay)changes state.The Allowable Values for the instrument settings are based on the RPS continuously providing a 56 Hz, 120 V t 10%(to all equipment), and 115 V+10 V (to scram and HSIV solenoids). The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels.The settings are calculated based on the loads on the buses and RPS HG set or alternate power supply being 120 VAC and 60 Hz.The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS bus powered components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass 1E source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring assemblies is required when the RPS bus powered components are required to be OPERABLE.This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1, 2, and 3;and in MODES 4 and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies (a control rod withdrawn in HODE 4 is only allowed by Special Operations LCO 3.10.-4,"Single Control Rod Withdrawal -Cold Shutdown").ACTIONS A.1 If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly (continued) BFN-UNIT 2 B 3.3-198 Amendment Cl RPS Electric Power Monitoring B 3.3.8.2 BASES ACTIONS (continued) will still provide protection to the RPS bus powered components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours)is allowed to restore the inoperable assembly to OPERABLE status.If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply(s)must be removed from service (Required Action A.l).This places the RPS bus in a safe condition. An alternate power supply with OPERABLE.power monitoring assemblies may then be used to power the RPS bus.The 72 hour Completion Time takes into account the remaining OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection occurring during this period.It allows time for plant operations personnel to take corrective actions or to place the plant in the required condition in an orderly manner and without challenging plant systems.Alternately, if it is not desired to remove the power supply from service (e.g., as in the case where removing the power supply(s)from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken.B.1 If both power monitoring assemblies for an inservice power supply (MG set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost.In this condition, 1 hour is allowed to restore one assembly to OPERABLE status for each inservice power supply.If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status,, the associated power supply(s)must be removed from service within 1 hour (Required Action B.l).An alternate.power supply with OPERABLE assemblies may then be used to power one RPS bus.The 1 hour Completion Time is sufficient for the plant operations personnel to take corrective actions and is acceptable because it minimizes risk while allowing time for (continued) BFN-UNIT 2 B 3.3-199 AMENDMENT 4l RPS Electric Power Monitoring B 3.3.8.2 BASES ACTIONS B.l (continued) restoration or removal from service of the electric power monitoring assemblies. Alternately, if it is not desired to remove the power supply(s)from service (e.g., as in the case where removing the power supply(s)from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken.'C.l and C.2 If any Required Action and associated Completion Time of Condition A or B are not met in NODE 1, 2, or 3, a plant shutdown must be performed. This places the plant in a condition where minimal equipment, powered through the inoperable RPS electric power monitoring assembly(s), is required and ensures that the safety function of the RPS (e.g., scram of control rods)is not required.The plant shutdown is accomplished by placing the plant in MODE 3 within 12'ours and in NODE 4 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.D.l If any Required Action and associated Completion Time of Condition A or B are not met in NODE 4 or 5, with any control rod withdrawn from a core cell containing one or more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action D.l results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods)is not required.BFN-UNIT 2 B 3.3-200 (continued) ANENDHENT Il 0 RPS Electric Power Monitoring B 3.3.8.2 BASES (continued) SURVEILLANCE RE(UIREMENTS SR 3.3.8.2.1 A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the entire channel will perform the intended function.Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is in a condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. The 184 day Frequency and the Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref.2).SR 3.3.8.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based on the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.SR 3.3.8.2.3 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual)signal, the logic of the system will automatically trip open the associated power monitoring assembly.Only one signal per power monitoring assembly is required to be tested.This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function.The system (continued) BFN-UNIT 2 B 3.3-201 Amendment J.m V'ho i, ib RPS Electric Power Honitoring B 3.3.8.2 BASES SURVEILLANCE REQUIREHENTS.S.(ti d)functional test of the Class 1E contactors is included as part of this test to provide complete testing of the safety function.If the contactors are incapable of'perating, the associated electric power monitoring assembly would be inoperable. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned, transient if the Surveillance were performed with the reactor at power.Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. 0 REFERENCES 1.FSAR, Section 7.2.3.2.2.NRC Generic Letter 91-09,"Hodification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System." 3.NRC No.93-102,"Final Policy Statement on Technical Specification Improvements," July 23, 1993.BFN-UNIT 2 B 3.3-202 AHENDMENT 0-II Enclosure V Volume 16 7Z~KM~SBIEIE VA II IL IEVA.IUTIHIGRITY .o UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OP~ Wc c(cfinA4 R<~oi Rv'>~~~o(K4~in 44('4'iiMl QfKanf a~c<(Pi'~bL~~>~cic Tt+c<',ca($ c;~+A'I I MSE AMP RIVuCRVXw4 1 J7 z Ani'AS A reasonabl Odo in se ous sub]act o safe limits ar liaits be ov vhfch the ce of e cladd and pr sist are eading ch a 1 t repair md,t ah dovn reviev Energy ocaf ssi before re tion o unit ration au a limit not in tself esult c Qolncos t it catos an orati dof ic Oncet regal ofay re@i liait safe aystea sett are se ings on tom ation initi te the utoaat protec Te acti at a 1 el such t the afc~iilits ll not.meed.The egion be eon the ety 1 t those A2(ae tings r resents gin no oyerati lying lov ese set.aarg hes be establi ed so vith oper oyera on of inst tati the sa ty liaits vill a er be moodedo C.-The 1 ting c tions for 0 ation s ocitf acceptab leva of ays ea rfo 4 noc~to~sa st and 0 ation f tho f ili>.Mm~condi iona are t, the lant'be o Orated sa and beozaal ituati can be elg c troll Za the eront a Lialt Condition r Operat and/or associa reysiroaen cannot bo isfied b use of cir~t os in moss f those 4ddr od in syecificat i the unit 1 bo ylac in at le t Hot tan@by vi 6 hours in Cold Shut vi the i~Big 30 uLloss c octiTo os aro c otod that rsLLt oporat under tho o?%isaib disc erg or ant the reacto is ylaced an oyera onal coadit on in vhich the syecifi tion is not yylicabl Ezeeyti to these epdreaeats 1 be st ed in th ivi syecificat, This oxides act to be A t for raastences t direct provided r in the 4 ficati and vhere ccurrence d viola the int t ot tho syecificati For lo, if 4 ication ls for tvo tees (or subset to be oyer~and yr ides for exp cit re~ts if e (or subarea ea)is inoye ble, thon t both terna r teas)ar inoperable aait is be in 1 t Ho Standby 6 hours and Cold Shu ovn vi foll 30 hour if the inoy>>ble condi on is n t co ctodo BFI Unit 1 1.0-1 QPP+yZ)~+On I~I NOV 18 1988 2.Cn a system, subsystem, train, componcn, or evice s determined to be inoperable solely because its onsite povcr aourc ia inoyerable, or sol ly because it offaitc power source ia inop rable, it may be co dered operabl for the purpose of satisfy thc requirements o ita apylicabl Limit Condition For Oyera on, provided: (1)its co sponding offsite or ieael yover s urce i oyerable;(2)all of its red t system(i)aubsy tern(s), train(a)f c cnt(s)f and device s)arc operab , or 1 evisc atisfy these quircments. Unlca both conditi (1)d (2)e satisfied, e unit shall be pl cd in at less Hot S dby 6 hours, in at least Cold tdovn vi the fol 30 hour This definition s not apylica le in C d Shutd vn or Refuel.Ttd.s provision describea aha additi nal condit must bc tiafied to permit peration to c tinue consist t vith the a ecificationa for r sources,.en an offaite onaite povc source ia not op ble.It ay cifically prohibits peration vh'ne division ia perablc bec use its offaite or ieael povcr ce ia inoyerab e and a syst subsystcmf t ain, compon, or device in other divisi is inoyerable fo another re n.This proviai permits th cquircmcnts aociatcd vi individual syst f subsystems f t ins, compon ta, or dcvic to be consist vith the r reacnta of e asaociat electrical pove source.It all operation o be govern by the time 1 t of the requi cats assoc ted vith th Limiting Condit on For Oyerat for the o site or die 1 pover source, not thc individ rcquirca a for each temf aubsystcmf train, coaponcntf or device t is determined to be inoycrablc solely because of the inoycrability of its offaite or diesel over source~D.Prior o ra th+fi t c trolgrod for th~c'hc cact criMcal'+"~Sr bA ldh8~k sy t f au syatcmf component, or cv cc oesUko perforating ics spec~fane on(s).all necessary attendant instrumentation, controls, noaaa3Aaaf emergency electrical ove~Casa~cool e&eal vatcr, lubrication,&other auxiliary equipment that arc required for the syatcmf subsystem, 4'~componentf or device to perform its function(s) are also capa e o performing their rclatcd support function s.5 t'ec'Q eJ esa~~F.Ope at me tha a tern co onen p o~ing ts t cd unc ons it rc re manne diate cans t th rcqu red a tion ill bc init tcd soo as pr cticab co der the afe era+oh, o e un t thc ortan of rc red ction.A o BFH Unit 1 1.0-2 AMENOQEIP vn.y 5 8

IIIII~~~.~~~~~I+~y~~~~I';~'ll~I~~~~~~~~~..i+tllj I itl~i'~~I~~.:e I'4I~I~~~~~~~'I~~~~~~~~I';~I'll i~SIAill'.~ill~i~~;~'llI'I I~DI~~~~I':~~~~~~'ll'~I~~~~~Ill~~~~I~'ll;ll crt~II~'I'I~~~~~I I','~I~~~'I~~~~~~~~~~~~'&0 A~~~~~~~~~~~~~~I~4~~~~I';~~~~JO~~~~~'~~~~~4'I II'~~"I~'wol tlVtll~i I all I I I AI~~~~~~~~;I'I~~~~~~~~'l I~'l l I I I t[IJgl~i~I if I I I~'I'~'we~~~~~~~I i'I l~CP~~~~~~;II I~~'ll~I ill 0~i~~~I~I I I'we~~~~~~I~'~~~~~~~~~ Q i cX';'on ae s c tres M.-The re ctor m de switch pos tion de ermines the de of ration f the r ctor en the e is f 1 in c re ctor v sel, ex pt that the Mod of Ope ation y rema n unc ged when th reactor de swi h is t porari move to ano her po iti as pe tted by he notes When here i no fu in th reac r vesse , the re tor is onsider d not be in any Mo e of ra ion or ope tional c dition.The r ctor de swi ch may then in any s tion or y be in erable.I 30-The reactor is in t e TARTUP/HOT ST Y MODE when the rea tor mode witch is the"STARTUP/H SThHDBY" osition.This is ften re rred to as just the STARTUP MODE.-The reacto is in the Run Mod<<hen the reactor mode sw tch is in the un" position.-The react r is in the Shutd Mode w r mode switch is in the"Shutdown" osition.era@c~.e e tor i+th~Reheel M+e a o de w tch is in the"34,fu+" poMtioh.See VuSQficwhen +~~~S foi Bi'm l$75 Sc~o~g)g The reactor mode switch may be placed in any position to perform required tests or maintenance authorised by the shift operations supervisor, provided that the control rods axe verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.~The reactor mode switch may be placed in the"Refuel" position while a single contxol rod drive is being removed from the reactor pressure vessel per Specification 3.10.A.5 provided that reactor coolant temperature is equal to or less than 212'.~The reactor mode switch may be placed in the"Refuel" position awhile a single control rod is being recoupled oz withdrawn provided that the one-rod-out terlock is OPERABLE.The reactor mode switch may e p ace n the"Startup/Hot Standby" position and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality. +<xug+AGLpon 4oe CW+cb Ska l5zs 3,'3~%,1 BPN Unit 1 1.0-4 ANENDMBrr 80.Z 9 6 0 1 RP III c ear>r~aŽ~ac/ze mwf a pover e ers to opera on t a react r po r of 3p 3 i5ftp this is also te ed 100 p cent pove and is t e po r leve authori d by th operat license.Rated st am f v, po rat coolan flov, tcd neut flm, rated clear tern ll~pres e ters to tho lees of ese psr stere eh the re tor s at rated pover.0-Pr ry cont ament in cgrity means that the ryvcll and pr sure suppre sion cr are i act and all:..-of the fol ving conditio are satis cd: 1.All nonau matic containm t isolati valves lines c ected to the reac r coolant sys or cont ament v ch arc n',f35 quired to b open during a ident con tions ar closed, ccpt fo valves tha are open under administra ve contr as pc tted by Spe ficatioa.3.7. kt lea t one door each airlock i closed scaled.1 auto tic contaiam t isolatioa val s are 0 RABLE or each 1 vhich ontains an i pcrable isolati valve isolated as requ ed by ecification 7.D.2.4.kl bli flanges and manvays are closed.Secondary tainmeat integr means that required unit eactor zone and refueling z are intact the folloving c itions are ct: a)least one r ia each access pcning, to the urbine bu ding, coatro bay and out-of-d rs is closed.b)The s by gas tres cat system is 0 and can aintain 0.25 in es of vater n ative pressure those areas herc sec coataizslent ia grity is stated 0 exist~c)1'seconds contaiament p tratioas requir to be clos accidcn conditions are ither: 1.Ca ble of be closed by aa RhBLE seconda cont ent aut atic isolation tcm, or 2.Closed at least e secondary con inment autom tic solation alve deact vatcd in the iso ted positio.2.actor zon seconds contagium integrity means thc unit re tor build is in ct and thc olloving condit ns arc met: BFH a)ht cast one or bct en any op to the turbin buil ng, contr bay d out-ofdoo is closed.'AQE&OF~em>>~to,r 97 .l 0 P.S co 5p~',g;ca dijon I (c e a)NOV 18 1988 2.b)The stand gas treatment stem is OPE and-can maintain p.25 inche water negative p essure on the it xone.c)All the unit eactor building v tilation sys penetrations required to be losed during acc ent conditio are either:.1.Capable of be closed by an 0 RABLE reacto building ventilation sys em automatic iso tion system, r g3 4 2 Closed by at leas one reactor buil ng ventilati system automatic is lation valve dea ivated in th solated position.If it is de irable for operatio 1 consideratio a reactor one may be isola d from the other r ctor xones and e refuel zo e y maintaining at least one closed oor in each co on passage y b veen zones.*Reactor xone safet related feature are not corn romised by o ings betveen adjac t units or ref el zone, unle it is desir d to isolate a giv zone.Refuel ne seconda containment integri means the re el'zone is intact the foll vtng conditions are et: a At least ne door in ach access opening, the out-of-d ors is closed.b)standby g s treatmen system is OPERABLE can mainta 0.inches va r negative pressure on the ref 1 xone.c)All re uel xone v tilation stem penetrations r quired to be clos during ac dent cond tions are either: 1.Capabl of being cl sed by an PERABLE refuel zo ventilat on system a omatic i lation system, or 2.Closed by a least one fuel xone ventilation syst automatic is ation valve eactivat in the isolated position.If it is esirable for op rational co ideratio the refuel one may b isolated from e reactor x es by mai taining all tches in ace betveen the refuel floo and react zones and at east one osed door in e access be veen the r fuel zone and t e reactor uilding.*Re el zone saf y-related eatures are no compromis d by openings tveen the r actor buil ing unless i is desir to isolate a ven xone.t*To e ectively c trol z e isolat n, all:accesse to the af cted zone vi11 be loc ed or guar ed to pr vent unco rolled passage to the una ected zones.BFH Unit 1 1.0-6 AMENOMENT~ y g 8 ~Qjl<C<fiCcfli jH//-Interra1 be the cnd of one refuel%outage f a gait, and the end of c ext subsequent refuel outage r the same unit$38 S~We 4/k'~+~c t t<q Case.<qCTHh~I shutdovn of the unit prior to a refueling, and the startup of the unit afte that refueling.

For the p se of designating equency of testi and surreillance, a refueling utage shall mean egularly schedul utageg hoverer, vhere such o es occur vithin months of the coaplet of the prerious refueling outage, the repaired surreillance testing need not be erforsed next regularly ut/OC+pc~I~g w<<r<*~KJUSRCRS-1 sources~rcactirity control c~poaents, vi the reactor reseal vith the reseal head rcaored the ress Sevcaaat of source range aoaitors, teraediate range rs, rarersing in-core probes, or special so@able detectors (including underresael replacement~'P coapl ioa of aoreacat of a c eat t a safe S<Vwl.~m qc 0 ii FAt Itic/<<mq t s 0 cat y+eac or reseal pres listed ia the Techai Specifications are se measured by the react reseal steam s ce detectors SLdt bc S~a/kC Cc~g~~~4 fbi'c<W Qual>ritical pover ratio i is calculat to cause arne point ia the asseably to exper ence boiliag ransi oa act asseRbly operating pover o cia q<r~6 L boiling rcg b nucleate and fila iag.Traasitioa bo is the regime vhich both nucleate a~Ha boiliag occur iateraitteat vith neither type being coayletely stabl 3e t ra ass/>~~4~t~g the re>of the fuel rod polrer girea asseably and ocatioa to/,'~if<~pcsew ensity/ft)at that location~C~.applicable height and is equal to the nm of the for all the fuel rods ia the specifi height dirided by the amber of fue rods-The 61'.pJ)ocat ons ity (M for a liait fue Apgcz to a specific planar e at the specified ia the fuel bundle.D~e.ZQ CF l

Qp))5.-GO MhXIMUM I OF CRITICAL P R is the aaxianm v ue of the etio of the flo orrected CPR op ting 1lait found the CORE OPgggIgC TS REPORT divided by the actual CPR for all uel lies in the cor V.lo 2o-kn instrmeat calibration ILeans the ustILent of an instrLaent signal outpat ao that J.t co esyonds f vf thiIL acceptable range, and accuracy y to a)QloRL vela (s)of the paraaeter ch the instant tora.@gals+-A channel is an arr t of the sensor(s associat caapceeats used to ev te ylaat variables yroduce di ete outyata used la lo c.A charnel termlna s aad loses i identity Mere individ chaImel outputs are caabiaed ia 1 lc.3~claw+;g;~. ~a~a~e~)~+~S4etkg 5~sl-Aa iastnaNat ctional test means the hQection of a iILulated signal into the trmaent priLnary seasor to verify the proper lastnment channel response, alarIL and/or initiating action am+~~k ke qualitative by observation of behavior daring oyeratim.This deterainatioa shall include, where possible, coayarisoa of the ependeat lnsttlsleats easurlng the sess v4%44ble tt v)l'et.a test of al relays asa o)s)teats f a lo io oirosit pq~se s'w ac~I.o-gP go):'~~~"~)ztuda charnel trly signals ead aaxfliary equiyeeat required to iaitlate actloa to accoaplish a protective triy fanctioa.rip syatca say require oae or aors lastnaent channel trip related to cme or re plant yaraaeters la order to tiate triy systea action.Zaitiatioa of yrotect e action aay squire the trlpylng of a e trly systea or coiac dent tripping of two trip teas.-kn action laltla ed by the protectioa sy tern vhea a 1 ls reached.i yrotectlve ction caa be at a charnel or tea level.-A system protective actloa vhich results free the protecti e action of the channels aoaitoriag a yartlcular plant coaditioa. BT%Unit 1 NBSMBA'NO. 2 I B (3....

, from as close to the sensor as practicable up to, but not inc1uding, the actuated device, to verify OPERABILITY. The lOGIC SYSTEM FUNCTIONL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

gl 1.s"~1~+5 ftc4img'on I.I FES 2 4<ss5 9.d utoma c ctua-Simulated automatic actuation means plying a simu ated signa to the senso to actuate the circuit in q stion.10.-A logic i an arrangem t of relays, contacts, d other comp ents that pr uces a deci on output.(a)t-A 1 ic that rece ves signals rom channe s and pro ces decision utputs to th actuation 1 gic.~Q~NHG.AQ~cAbLs~acne r as of (b)-A logic t receives ignals (ei er from initiation logic or els)and p duces deci ion outpu s to complish a protective action 8/or ip the sensor o verif OPERABI TY incl ing al f ctions.fsyQAi n+AC)hall ba the adzes t, as necessary, of the channel out u u that it responds~necessary range and accuracy to knom values of the parameter~&eh t e channe monitors T.he olnmnal 99ii~ipfpn shall encompass the entire choate@inc u a amn and~trip functionepnd shall include the o rmt.The s~hasae naiihnLtSpn may be perfomced by any series of se uen al, overlapping or total channel steps t re channel is calibrated. on-ca ra a e.om ments aha a s r uirement, hut vent)be incluaed 4a channel functional west and sources che R/Will 7Psf LI-Shall be~rT.autre l l the~ecti of a simulated signal into the channel as close to e s or as practicable to~g<<~verify OPERABILITY including alarm emb4n trip functio@;~~hops atMNB RbvcnwnL%sr mes gcrfofrncd Q~o999 pf Rtysc<ici o+~a<>OVu>rgfnsS an m~4anrba>tCPSSP~fag CnHTC n CI.fS QSSCbL b 13 QhQ~Q AGE~~3 BFH Unit 1 1.0-9 NIlEHDMKMf NO.2 1 6 i(g l 3 i'72 SPCCil~cRh'aa l, I (~p A func" ional test is the manua operation or initiation of a sy em, subsystem, or omponent to rify chai'c functions within des'olerances (e.g.the manual ari, of a core pray pump to verify at it runs and tha it pumps the equired v lume of water).-The reactor is x a shutdown condit n when the r ctor mode s itch is in the shutdo mode position and o core alter ions are bei performed. An engineer safeguard is a sa ty system the actions of w ich are essential to a fety action requir d in response to a idents.~8<<5-A reportable event shal be any of those co ditions specifx d in Section 50.73 to 10 Part 50.BB.CC DD FF Am by Shall cont'n the methodol and paramete s used in the calculation o of f site doses resulting rom radioactiv gaseous and liquid ef fluent in the calculation f gaseous and iquid effluent monitoring Al/Trip etpoints, an in the conduc of the Environmental Radiol ical M itoring Pr am.The ODCM hall also contain (1)the Ra ioactive Ef uent Control and Radiologi al Environmental Monitoring rograms requ ed by Secti 6.8.4 and (2 descriptions of the informa on that s ould be inclgded in the ual Radiological Environmenta Operati and AnnualQadioactive E luent Release Reports requir Specific ions 6.9.1.S'nd 6.9.1.8.-The co trolled proc s of discharging air or gas rom the pri ry containment to maintai temperature, pressure, h midity, conc tration, or@ther operati g condition in such a ma er that replacement air ox gas is requ ed to purify the con inment.Qggjgg-e controlled process of di harging air gas from che primary conthinment to maintain temperat re, pressure, umidity,'I concentration~ or other operating conditi n in such a ma er cha"-replacement air or gas is not rovided or quired.Vent, used system names, d s not imply a nting proce s.BFN Unit 1 1.0-10 PAGE MS 373 5 ccrc gi ccgAA t/-Shall be that ine beyond which t land is not, , or otherwise con oiled by~.-Any area at or be nd the SITB BO to which ccess is nat contr led by the Licensee or purposes of pro tion o individuals from sure to radiation radioactive materials or y area within the TS BOUNDARY used for dustrial, ccamercial,. insti tional, or recreat onal purposes.miCa'OC<<aa jr'~DOSS BQUIVALRNT -131 s 1 be the concentration of I-131 He wage)alone would produce the same thyroid dose as the quantity and isotopic mixture of Z-131, Z-132, I-133, Z-13i, and I-135 actually present.The thyroid dose conversion. facto(sSsed for this calculation shall be those listed in Table ZZZ of TZD-1ilii~Calculation of Distance Pactors for Power and Test Reactor Sites-The charcoal adsorber seals installed on the discharge of the st)et air e)ector provide lay to a unit's offgas activity prior o release.-An individual in a Bo~rer, an vidual is not a any riod in which t indivMual receives defin in 10 CPR 20).trolled or CTED Ot TBR PIIC occupational dose (as-stzaaallat aattitaaatta shall ta t dutata tha OPKATI CCSDITIONS or'anditians specifi for individual limiting tions for ape tian unless otheieiso s ted in an individual illanco Requi eaeats.Bach Surveill ce Requirement shall be per rmed within the if ied surveillance erval with a maximua all le extension not to exceed 25 percent of specified illance erval.It is no intended that this (ext ian)pr isice be us repeatedly as a convenience ta extend illance inte s beyond t specified fa surveillances that are t perfo duriay r ueling outages.Porto of a illance Requir t within the specified ime iaCerval 1 constit te compliance OPRRAIILITY requirement f or a QaLtiay tian fo aperation and a iated action statements cosa a so requir by these specif tions.Surveillance WWI*I If it is discave ed that a s eillance was performed within;=s specified frequen, then c iance with the rement to dec.'a:e the XO not mot ma be delayed, frcwI the time of discovery, up"o 4 hours ar up to the 1 mit of the specified frequency, whichever s less.This delay period is permitted to allow performance of.".e surveillance. BPS 1.0-11 QAz~Zf the surv'llance is not performed within the delay perio the LCD must iamediat ly be declared not met, d the applicable con'tion(s)must be entere en the surveill ce is performed within e delay period and t rveillance is no met, the LCO must immedi tely be declared not me, and the applic le condition(s) must be tered.MM.S illance Requireme ts for ASME Section XZ Pu and Valve Program'Su eillance Requirem ts for Znservice Testing f ASME Code Class 1, 2, d 3 components s ll be applicable as folio s: Z rvice testing of ME Code Class 1, 2, and 3 umps and val s shall be perform d in accordance with Sect n XZ of the ASME iler and Pressure Code and applicable Adden as requir d by 10 CFR 50, Se tion 50.55a(g), except whe e specific written elief has been gr ted by the Coaeission pur uant to 10 CFR 5 Section 50.55(g))(i).2.Surveillanc intervals specifi in Section XZ of the AS iler and P essure Vessel Code d applicable Addenda for he i ervice tes'ng activities requ ed by the ASME Boiler Pre sure Vesse Code and applicabl Addenda shall be applicab as f llows in t se technical speci cations: ASME iler and essure Vessel Re ired frequencies Code appl icabl Addenda for erforming inservice termino for ins rvice~~ekly Mo thly Quarterly or very 3 mont emiannually o every 6 mon Every 9 mo hs Yearly or ually At least nce per 7 days At least o e per 31 days At least on per 92 days At least once er 184 days At least once r 276 days At least once pe 366 days 3.The ovisions of Sp ification'..LL are applicable t the above equired freque ies for perf ing inservice test ng activi es.Performan e of the above'nservice testi activities shall in additio to other speci ied surveillan requirements. 5.othing in th ASME Boiler an Pressure Vess Code shall be c strued to s ersede the re irements of an technical spe'fication. 6.The ins rvice inspe tion program f piping identi ied in NRC Generic tter 88-01 hall be perfo ed in accordan with=;".o staff posi ions on sch dule, methods, ersonnel, and mple expansion i luded in t's generic let r.BFN Unit 1 1.0-12 fthm OO.be a attern vh operat on a 1 s~~~"i)FE82419SS Pcc.iC'c towmcwr~-The COLR is the unit-specific document that provides for the cgrrent determined for each~ay cycle in accordance vith Specification Plant operation vi i e limits is addressed in ind v dual specifications. AC)LIMITI CONTROL OD Ph Rlf s+ll res ts in th core bei on a ermal imit,gi.e. ting lue for CRf f or R.A)3 gnseRY)Cz pages)O 7-n/SEg, 7'gi ice>A Jg msea7'Cai fa~cc)BPS Unit 1 1.0-12a atm0mr S0.2 Z 6 Qf Q i lNSERT1'Page 1 of 3)Term ACTIONS LEAKAGE Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated conditions within specified Completion Times.LEAKAGE shall be: a.Identified LEAKAGE LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank;or 2.LEAKAGE into the drywell 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;b.C.d.Unidentified LEAKAGE All Leakage into the drywell that is not identified LEAKAGE;Total LEAKAGE Sum of the identified and unidentified LEAKAGE;Pressure Boundar LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System{RCS)component body, pipe wall, or vessel wall.The definitions found in this insert will be placed in alphabetical order with the other ISTS definitions. INSERT 1 (Page 2 of 3)LINEAR HEAT GENERATION RATE (LHGR)NODE The LHGR shall be the heat generation rate per unit length of fuel rod.It is the integral of the heat flux over the heat transfer area associated with the unit length.A NODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.PHYSICS TESTS SHUTDOMN NARGIN (SDN)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 13.10, Refueling Test Program, of the FSAR;b.Authorized under the provisions of 10 CFR 50.59;or c.Otherwise approved by the Nuclear Regulatory Commission. SDN shall be the amount of reactivity by which the reactor's subcritical or would be subcritical assuming that: a~b.c~The reactor is xenon free;The moderator temperature is 68'F;and All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. lith control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDN. 0 INSERT 1 (Page 3 of 3)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.TURBINE BYPASS SYSTEM RESPONSE TIME The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components: a 0 b.The time from initial movement of the main turbine stop valve or control valve until 8N of the turbine bypass capacity is established; and The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.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.PAGE/oF K> , INSERT 2 Definitions

1.1 Table

1.1-1 (page 1 of 1)MODES NODE TITLE REACTOR NODE SMITCH POSITION AVERAGE REACTOR COOLANT TEHPERATURE ('F)Power Operation Startup Hot Shutdown(a) Cold Shutdown(a) Refueling(b) Run Refuel(a)or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel NA NA>212<212 NA (a)All reactor vessel head closure bolts fully tensioned.(b)One or more reactor vessel head closure bolts less than fully tensioned. BFN-UNIT 2 1.1-7 Amendm/0 QF 0 Il4SERT 3 (Page 1 of 21)Logical Connectors 1.2 1.0 USE ANO APPLICATION

1.2 Logical

Connectors PURPOSE The pur pose 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~A and QB.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 indentions of the logical connectors. @hen 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 Thc following examples illustrate the use of logical connectors.(continued) BFN-UNIT 1 1.2-1 Amendment 0 INSERT 3 (Page 2 ot 21)Logical Connectors 1.2 1.2 Logical Connectors EXAHPLES (continued) AHP ACTIONS CONDITION REQUIRED ACTION COHPLET ION TIHE A.LCO not met.A.l Verify..'~AN A.2 Restore...In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.l and A.2 must be completed. BFN-UNIT I 1.2-2 (continued) o~~Amendment lNSERT 3 (Page 3 af 21)Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES (continued) ACTIONS CONDITION RE(VIREO ACTION COMPLETION TIME A.LCO not met.A.l OR Trip...A.2.1 Verify...A.2.2.1 Reduce...A.2.2.2 Perform...A.3 Align...This example represents a more complicated use of logical connectors. Required Actions A.I, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector+0 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+0 indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed. BFN-UNIT 1 1.2-3 Amendment p,~p 2 P.OF~+

lNSERT 3 (Page 4 ot 21)Completion Times 1.3 1.0 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 Times(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 NODE 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 divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will~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.(continued) BFN-UNIT 1 1.3-1 Amendment FAGS P8 OF~~ INSERT 3 (Page 5 of Completion Tines 1.3 1.3 Completion Times DESCRIPTION (continued) H, t~bdIdl, hyt, p t, 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~f r~inoperability; and 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 division, 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"frow discovery..." Example 1.3-3 illustrates one use of this type of Completion Time.The 10 day Completion Time specified for Condition A and B in Example 1.3-3 may not be extended.BFN-UNIT 1 1.3-2 (continued) Amendment z4 nt-F~ lNSERT 3 (Page'6 et 21)Completion Times 1.3 1.3 Completion Times (continued) EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.Required Action and associated Completion Time not met.B.l Be in HODE 3.~AN B.2 Be in HODE 4.12 hours 36 hours 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 HODE 3 within 12 hours gg in HODE 4 within 36 hours.A total of 12 hours is allowed for reaching HODE 3 and a total of 36 hours (not 48 hours)is allowed for reaching HODE 4 from the time that Condition B was entered.If HODE 3 is reached within 6 hours, the time allowed for reaching MODE 4 is the next 30 hours because the total time allowed for reaching HODE 4 is 36 hours.If Condition B is entered while in MODE 3, the time allowed for reaching HODE 4 is the next 36 hours.(continued) BFN-UNIT 1 1.3-3 Amendment OF FE 0

1.3 Completion

Times INSERT 3 (Page 7 of 1)Completion Times 1.3 EXAMPLES (continued) AMP.3-ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.One pump inoperable. A.l Restore pump to OPERABLE status.7 days B.Required Action and associated Completion Time not met.B.l Be in MODE 3.~AN B.2 Be in MODE 4.12 hours 36 hours When a pump is declared inoperable, Condition A is entered.If the'ump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.l and 8.2 start.If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and B are'exited, and therefore, the Required Actions of Condition B may be terminated. 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.(continued) BFN-UNIT 1 1.3-4 Amendment~(rw YS

1.3 Completion

Times INSERT 3 (Page 8 of 2 Completion Times 1.3 EXAMPLES.2 (ti d)awhile in LCO 3.0.3, if one oF the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expir ed, 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 i.f 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.(continued) BFN-UNIT 1 1.3-5 Amendment pAGp~7 0

1.3 Completion

Times INSERT 3 (Page 9 of 21)Completion Times 1.3 EXAMPLES (continued) P.3-3 ACTIONS CONDITION REQUIRED ACTION.COHPLETION TINE A.One Function X subsystem inoperable. A.l Restore Function X subsystem to OPERABLE status.7 days 10 days from discovery of failure to meet the LCO B.One Function Y subsystem inoperable. B.l Restore Function Y subsystem to OPERABLE status.72 hours 10 days from discovery of failure to meet the LCO C.One Function X subsystem inoperable. One Function Y subsystem inoperable. C.l Restore Function X subsystem to OPERABLE status.+0 C.2 Restore Function Y subsystem to OPERABLE status.12 hours 12 hours (continued) BFN-UNIT 1 1.3-6 Amendment PAGE~go F INSERT 3 (Page 10 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES l.3-3 (tl d)Rhen one Function X subsystem and one Function Y subsystem are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and" Condition B are tracked separately for each subsystea, starting from the time each subsystem was declared inoperable and the Condition was entered.A separate Completion Time is established for Condition C and tracked from the time the second subsystem was declared inonerable {i.e., the time the situation described in Condition C was discovered). If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are'xited. If the Completion Time for Required Action A.l 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 subsystem was declared inoperable {i.e., initial entry into Condition A).The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met.In this example, without the separate Completion Time, it would be 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.The separate Completion Time modified by the phrase"from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO.This Completion Time allows for an exception to the normal"time zero" for beginning the Completion Time"clock".In this instance, the Completion Time"time zero" is specified as coaeencing at the time the LCO was'nitially not met, instead of at the time the associated Condition was entered.{continued) BFN-UNIT 1 1.3-7 Amendment PAGE A~DF~~

INSERT 3 (Page 11 of 21)Completion Times 1.3 1.3 Completion Times EXAHPLES (continued) '.-4 ACTIONS CONDITION REQUIRED ACTION COHPLETION T!HE A.One or more A.1 Restore valve(s)4 hours valves to OPERABLE inoperable. status.B.Required Action and associated Completion Time not met.8.1 Be in HODE 3.AND B.2 Be in HODE 4.12 hours 36 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.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 hour s (plus the extensions) expires while one or more valves are still inoperable, Condition B is entered.(continued) BFN-UNIT 1 1.3-8 Amendment pygmy 90 Ot'~~~

INSERT 3 (Page 12 ot 21)Completion Times 1.3 1.3 Completion Times EXAMPLE (continued) P.3-ACTIONS NOTE-Separate Condition entry is allowed for each inoperable valve.CONDITION REQUIRED ACTION COMPLETION TIME A.One or more valves inoperable. A.l Restore valve to OPERABLE status.4 hours B.Required Action and associated Completion Time not met.8.1 Be in MODE 3.AND B.2 Be in MODE 4.12 hours 36 hours The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked.If this method'f 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.The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis.Qhen 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.(continued) BFN-UNIT 1 1.3-9 Amendment ne YC 0

1.3 Cbmpletion

Times INSERT 3 (Page 13 at 21 r Completion Times 1.3 EXAMPLES BSLPP.-5{ti dl 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'ondition A expire, Condition B is entered separately for each valve and separate Completion Times star t 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.AMP.-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.One channel inoperable. A.l Perform SR 3.x.x.x.Once per 8 hours A.2 Place channel in trip.8 hours B.Required Action and associated Completion Time not met.B.1 Be in MODE 3.12 hours (continued) BFN-UNIT 1 1.3-10 Amendment ,;..-..ga.o.-~5 0 INSERT 3 (Page 14 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES BNIRLEJ.3.6 (ti d)Entry into Condition A offers a choice between Required Action A.l or A.2.Required Action A.l 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.l begins when Condition A is entered and the initial performance of Required Action A.l 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.(continued) BFN-UNIT 1 1.3-11 Amendment p~Gp gS OF~~ 0 j' INSERT 3 (Page 15 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES (continued) P.-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.One subsystem inoperable. A.l Verify affected subsystem isolated.1 hour Once per 8 hours thereafter A.2 Restore subsystem to OPERABLE status.72 hours/B.Required Action and associated Completion Time not met.B.l Be in MODE 3.B.2 Be in MODE 4.12 hours 36 hours 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.l.If after Condition A is entered, Required Action A.l 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.l (continued) BFN-UNIT 1 1.3-12 Amendment par p PY

1.3 Completion

Times INSERT 3 (Page 16 of 21))Ig Completion Times 1.3 EXAHPLES BllKLJ 3 7..(ti d)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.IHHEDIATE When"Ittlediately" is used as a Completion Time, th>>COHPLETION TIHE Required Action should be pursued without delay and in a controlled manner.BFN-UNIT 1 1.3-13 Amendment ZF oFYE

1.0 USE AND APPLICATION

1.4 Frequency

INSERT 3 (Page 17 of 21)pi<)Frequency 1.4 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 Limiting Condition for Operation (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, 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.Example 1.4-4 discusses these special situations. 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 (continued) BFN-UNIT 1 1.4-1 Amendment INSERT 3 (Page 18 of 21 Frequency 1.4 1.4 Frequency DESCRIPTION (continued) criteria.SR 3.0.4 restrictions would not apply if both the following conditions are satisfied: a.The Surveillance is not required to be performed; and b.The Surveillance is not required to be met or, even if required to be met, is not known to be failed.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 I, 2, and 3.EXAMPL.4-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 interval specified in the 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 (continued) BFN-UNIT I 1.4-2 Amendment pggE g7 OF~

INSERT 3 (Page 19 ot 21 Frequency 1.4 1.4 Frequency EXAHPLES~4.(ti 4)otherwise modified (refer to Examples 1.4-3 and 1.4-4), 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 HODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the HODE or other specified condition. Failure to do so would result in a violation or SR 3.0.4.p.4-SURVEILLANCE REgUIREHENTS SURVEILLANCE FREQUENCY Verify flow is within limits.Once within 12 hours after)25%RTP~ND 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<25K 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 extension allowed by SR 3.0.2.(continued) BFN-UNIT 1 1.4-3 Amendment nr~~WR OF 0 INSERT 3{Page 20 of 21 Frequency 1.4 1.4 Frequency EXAMPLES~EAEP.4-)tf d)"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'X RTP.~EEAAP.4-4 SURVEILLANCE REgUIREHENTS SURVEILLANCE FRE(UENCY NOTE Not required to be performed until 12 hours after>25K RTP.Perform channel adjustment. 7 days The interval continues whether or not the unit operation is<25%RTP between performances. 4 th ht dfff th 44 4~I fth 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 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 HODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power>25%RTP.(continued) BFN-UNIT 1 1.4-4 Amendment E<<gr.g 5g OF~+ 0 lNSERT 3 (Page 24 of 24 Frequency 1.4 1.4 Fi equency EXAHPLES EMEILJ.4-(t'nce 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, there would then be'failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.p.¹-4 SURVEILLANCE REQUIREHENTS SURVEILLANCE FREQUENCY-NOTE-Only required to be met in MODE l.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 HODE 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 (plus the extension allowed by SR 3.0.2)interval, but the unit was not in HODE 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 HODES, even with the 24 hour Frequency exceeded, provided the HODE change was not made into HODE 1.Prior to entering HODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.BFN-UNIT 1 1.4-5 Amendment 0 Sfhcigaz o I.I NY 2 0SS3 THIS PACE ISTEHTIOKALLT B2%Uaie 1 i+0-X2b SPec'e NOV 22 1988 8 (ShcL)D (Dally)(Vee36y)(er!y)f}<I~<Sesa-(-early)R'efnelin)0/0 (St-Uy)kt least e yer 1 hoar kt Least onc yer no c endear 24 hoor Cay (ght to').kt east once yer 4ays.kt le once yer 31 ys.t least~yer 3 aea or 92 kt eaat once er 6 aumchs 184 Ca kt le t once ye year or 366 y3.it least e yer o ratina cycle PTlor to e reactor s tokyo t ayylicable. P.(or)Caay e4 ytior to ach release.BiK Unit 1 x.o-u AMENOMBT NO.$5 9 p~iaE~koF

THXS PACE ZÃaiTIONALLY L BLOK Unit l.0-l 3a AQMMgp gg gg 9

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP sDcci+co~i~I I N0TS h~c+esg41 gpw~g~g Ms gggflom epp)tet P'+I'~cP~/Pc-~c/+~spp Iic'q5Q~'8"~I>"chic Tcc 4~a'ciI spec,g~~pz Y4rw ,~.Agar safety fmits e lfmf bel vhi r le C e of cladd sad r sys ebs~assur.Ezc ch a 1 t re res t ahu dovn ev the toaf c gy ssi before seamy ion o uaf o ratf Oper fon b a 1 may C fn Ctse?d t in erf ous oasequ es t it catea oye ti de ci sub5 C to r gulato r 1fmf t sa ty stem seC seC inst ta on vhi fait te aut tic pr tact e act at level ch t the afe limi s ll be eed The region betve the s ety 1-t ch e s t repr ts gfn th ao op atf on fag lov ese e fags.The in be estab shed o thaC th per o r ion the t tati the ety ta ll er b-The Xfmf ing oadi ous for erat sy ffy e ceyt le 1 els f tea erfo~cess Co sar safe oyer Cion f th f flf thea c ti are, pl c be erat abao s ti be tro ed.the er t a L fng tion fo Oyerat aad/or as ciated eqafra ts c be sat fied b use of cf tsnc in exc of tho~sddres in syeci cation the ani shall b ylaced at le t Eot t vithfn 5 hours in Co Shat vf the 30 s anlesa correctf aeasur are c lated t p t oyer fon and the ye saibl disc ery o untf1 the react r is pla ed in an erat coadf on in ch syecf f tion f not ayyl able.Except to e rasn ahLL1 scaced the vf syec icatf.This rovid actions t be for circles ances t dirac proof ed for in f feat sad re carrenc voald late the int t of syecf catfon For le, i a ication calls f tvo tas (subsya erne)to be ab snd y sides exyl cit r raents f one ta r sub ta)is yer le, th if bo systems (o subsy as)e inoyer le uaft i to be at le Eot andby 5 and Cold tdovn thin the f 30 ho s if the inoyer le c tion i uot, correct BFS Unit 2 1.0-1 0 Qp, NOV 18 1988 en a syst, sub stem train compo ent, o devi e is d c cd t be in perab c sole beca se its onsit pov so ce i inop rablc or s lely b ause ts of ite p vcr s ur~s oper le, t may e co idcr opera e for thc p osc f$s is ing c r ir ts its a lies e L ting C adit n fo Op ati , pr ided:)its correspo ing of ite or cscl po cr so e is o able, and (2)all of ts redun t syst (s), s syst s), tr (s), omponen s), device(s arc op able, like sc sat fy the requi ents.Unless b cond ions (and)are s tisfi the t shall bc place ia at sst Ho Standb vithin 6 hours and in at less Cold Shu ovn vi n the folio 30 ho s.def ion is n appli able in Cold utdovn or Refu ing.s pro sion des ibes v t addi onal c itio must b satisfi d to pe t'opera ioa to ontinu co istcat vith th spccifi ations or povcr ourccs, vhen an offs te or ite po er sour is no operable It sp cifical y prohi ts op ation v cn one vision s inopcr lc bec se its offsit or di el pove source s inope able and syst bsyat tra , comp t, or cvice another vision is pcrabl for other r soa.s prov ion permi s thc req ircmeat asso iated vi indiv ual sy ems, sub stems, tra, comp eats or devi es to be consist t vith t"requi cments f the ssociat electr cal pov source.It allova perati to govern by the time 1 t of the requir ts as ciate vith th Limit Condit on for eratio for tne offsi or dic cl povc source, ot the ividual rcquir eats r each stem, ubsystem, train, corn ent, dcvic that dete ed to be inope ble solel beca e of inopc bili of its ffsite or diesel pover oourc~8.8 oaf'/ld cJ kLA g(yi'>G~system sub cmp compoacnt, or d~Hce shall be crab or have o e nbili vhea it is capable of performing its spec icd function(s gall cssary attendant instrumentation ntrols, no<emergen lectrical power j40Qpcos'cool seal vater, labriratiot@4~%ther auzflkaxT g>...~g cquipmcat that are required for the system, subsystem>> compoacat, or device to perform its function(s)-are able of performiag their rclatcd support function(s). ~ceil~~4+the".ra eri BFH Unit 2 1.0<<2'AMENDMENT go.y Sg ' 1 88 l S t.ri4t'C~<~ l t 1.0 (Cont'f<j'/rw~e w 0 o-Rea tor pover operation s any ope tio A6 in th BEBEEDP/H01 11'HDBX or DH MODE th the re tor eriti 1 d above percent rat d pover.0-e reactor in the ARTUP CO TIOH vhen the p@w thdrawal of control ods for th purpose making t reactor cr tical ha begun, rea tor power less th or equal to 1 perce t of ted, an the react is in the STARTUP/H STAHDBY ODE.0 S CO 0-reactor in the H STAHDBY OHDITIOH when r ctor po r is less han or eq al to 1 pe cent of ra ed.The reactor s in th STARTUP/I STAHDBY ODE, and t e reactor not the S TUP CO ITIOH.reactor oolant tern erature ma be g ater 212 A37 g/Hot that a OT STAHD COHDITI cannot ist simul eously vi h a STAR COHD IOH due the dif rence in ntent.A OT STAHDB COHDI OH exi s vhen t reactor ode swit is place in the STAR HOT ST BY posi on (for ample, t comply vi an LCO)and pov level as been duced to percent or lover.time control ds are eing vith rawn for he purpo e of incre sing actor p er leve, the rea tor mode vitch ha been plac d in the S/HO STAHDB position, and reac r pover evel is at or belov on percent, a STAR COHDIT H exists.0 CO 0-e reacto is in th SHUTDO COHDITIOH hen the r actor is n the S utdovn o Refuel M de.1.-0 S 0 0 0-The r ctor is n the HO SHUTDOWH CO TIOH vhen reactor oolant peratu is grea er than 212 and the eactor in the HUTDOWH COHDITIOH A, 0 0 e react r is in the COLD DOWH COHDITI vhen reac or cool t tempe ature i equal to r less than 212 F and the eactor in the SHUTDO COHDITION, CO CO 0 The react r is in he COLD COHDITI vhen reictor cool t temperat re is equa to or 1 ss than 212 F any Mode of Opera ion (except as defined in K.2 a ove).~', BFH Unit 2 1.0-3 AMENDMENT NO 15 4 pAGp~O~ M.5 eC.Wse A dP./R 30 1993-The reactor mode s~itch position determines the Mod of Operation of the rea tor when there is fuel in the reactor vesse , except that the Mode Operation may remain un anged when the re tor mode switch is tempo rily moved to another p ition as permitte by the notes.When ther is no fuel in the react vessel, th reactor is considered no to be in any Mode of Op ation or operati condition. The reactor mode switch may then be in!any position o may be inoperable 2~-The reactor s in the STARTUP/HOT STANDBY MODE when the reacto mode switch is in the"STARTUP/HO STANDBY" position.is is often ferred to as just he STARTUP MODE.-The reactor s in the Run Mode when the reactor mode witch is in the"R" position.3~-The reactor s in the Shutdown Mode when th~g)o node switch ds in the"Shutdown" posdtio 4~"""'"'"'I""'eactor mode switch is in the"Refuel" osition.~L 3 ed's~tcdiOW far C44+)Cthe+Sssl fttm Sech'o~s.to The reactor mode switch may be placed in any position to perform required tests or maintenance authorised by the shift operations supervisor, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.~The reactor mode switch may be placed in the"Refuel" position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.10.A.5 provided that reactor coolant temperature is equal to or less than 212'.~The reactor mode switch may be placed in the"Refuel" position while a single control rod is being recoupled or withdrawn provided hat the one-rod-out interlock is OPERABLE The reactor mode switch may be placed in the"Startup/Hot Standby" position and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the eactor to criticality. BFN Unit 2 S<<~wzQicqkja~ 4a~C~~~~~~<~~sm P.Z.g.~1.0&NENDMENTHO 2 y2'Ig. 0 gWsRaA~P~4~<cR4>o~.l.f<<P s4//bc a.4o&rc~ck r~4Co+f~~s4~ago W Qf 3 p+gal p"'~~o4+ef 3og3 cut: ow-ated ower refers o operation a a reactor ower of t.this is al termed 100 ercent power d is the imum po cr leve authorised the operat license.R ed steam low, cool flow, rate cutron flux, d rated nuc r syst pres re refe to the valu of these par eters when t reactor s at rat d power.0.Co e Primary cont inment inta rity cans that e drywell d pressure s pression ber are in ct, an all of the ollowing nditions are tisfied:/ISLES l.All n utomatic ontainment is lation valve on lines onnect to the eactor co ant systems o containment which are ot required o be open during accid t conditions axe closed except for valve that are en under adm istxative c troi as pcrmittcd b Specifica ion 3.7.D.t least one or in ca.airlock is osed and se led.3.Al automatic co tainment olation val s are OPERA in which conta an inopc able isolati n valve is qu red by Specif ation 3.7.2.or eac olated a 4.1 bl flanges and manways a closed.P~S Scc dary containmcnt tegrity me that the r quired un t react zones and refue zone are tact and t c followi condit ns are met: a)At lea t one door in ea access op ng to the t rhine build control bay and at-of-door is closed.(AQ b)The stand gas treatment tern is OPB LB and can aintain.25 inches water negative ressure in ose areas here condary con inment integrit is stated exist.c)All ccondary co ainment penctra ions requir d to be clo ed duri accident co itions are eit er: 1.Cap le of being osed by an OP LB sccon ry coats cnt automa c isolation s tern, or Closed at least on secondary co ainment aut atic isolation lvc dcacti ted in the i lated positi n.2.cactor zo e seconds containmcn integrity mc thc unit actor bui ing is int ct and the ollowing con itions are m BFH Unit 2 a)least on door bet b ding, co rol bay 1.0-5 any ope ng to thc t bine out-of-d rs is close F='=-~ax~AMENDMEMT gP.P Z g P.Seconda ta ent Inte r (Cont'd)NOV 18 1988 2.b)The st dby gas treatment system is OPE LE and can int 0.25 inc es water negative ressure on th unit zone.c)All the uni reactor building entilation sy em penetrat ns required to b closed during ac dent conditio s are either 1.Capable of b ng closed by an ERABLZ reacto building, ventilation sy em automatic iso tion system, Closed by at least one reactor buil ng ventilation ystem automatic isolation alve deactivated n the isolate osition.If it is de irable for operation considerations, reactor zone may be isola d from the other rea tor zones and the ref'uel zone by maintaini at least one closed or in each commo passageway between zones.*Reactor zone safety-elated features e not.compromised by o ings between ad)ace units or refuel zone, ess it is desi d to isolate a given one.3.Re el zone seconda containment integrit means the refue zone is i act and the foll wing conditions are m t: a)ht ast one door in ach access opening t the out-of-doors is cl ed.b)The Stan y Gas Treatmen System is OPERABLZ an can maintain 0.25 inch water negative ressure on the refue zone.All refuel z e ventilation stem penetrations req ired to be closed during ccident condit ns are either: 1.Capable of be closed by OPERABLE refuel zone entilation sy em automatic olation system, or 2.Clo ed by at leas one refuel z e ventilation system auto tic isolation valve deactiv ted in the isolated positi If it is desirable for operatio 1 considerat ons, the refuel zone y be isolated fro the reactor ones by main ining all hatches i place between the fuel floor d reactor zo es and at least one closed door in each access betw n the refuel one and the reac r building,.* Refu zone safet-related feat res are not compro sed by openings be een the re tor building ess it is desired o isolate a given ne.*T effe tively control z e isolation, all a cesses to t affected zo will be locke or guarded to pr ent uncontrolled ssage to th unaffected z es.NEHDMNT NO.I 5 4 BFH Unit 2 X.0-6[ 0 NAY" 0]g93~a C c-Interval be ecn thc end of onc r ueling outag for a pa cular unit and the en of the next subsequ'efueling gL41 outage for the s unit.S.ti,C+f Ceej~)Mp g cr W Eggs'.C~gsiyi gated: o-Refueling outage is thc period of time betveen the shutdown of the unit prior to a refueling and the startup of the un t after that refueling. or the purpose of designati frequency of.sting and surveillance, efueling outage shall m..an a rcgul ly scheduled outage;howe r, where such outagcs occur, vithin 8 mont of the completion of the previous refueling outage, the required urvcillance testing need not be performed until thc next regularly scheduled outage.~(3 teM S iv Vp tote sources, reactivity control components vithin the reactor vessel vith the vessel head removed and fuel in t e ves Movement of source range monitors, ntcrmed ate range mon tors, aversing in-core probes, or special ovable detectors (including undervesscl replacement)~-~~ ~.+Suspension of CORE hLTERATIOHS shall not reclude completion of'ovement of a component to a safe~fr g+~~p prki<rrC no Ogo$e~gas4P 4$SC 4 IIV I/md d$C'OC ra CC tt CCQ 4.3'~a o ess essu-Unless ot erw se ndicated, r actor vessel pressure sted in the Tcchnical S~~fications a asured b the reactor ssel steam space detectors. /1/s'4A Hics~/le C (CpR)4E.CWS/S i', W Cei<ritical povcr ratio which s calculated to cause some point in thc assembly to experience boiling ransition,~the actual assembly operating pover.~~licg,'c if vE~J;v', op~~aA4, Cupped.1m'g~(S)-Trans t n ngm reg me betvc nucleate and film boil Transition bc~as is the regime in ich both nucleate and boiling occur intermittently with neither type being completely stabl QAn BFH Unit 2 3~a 0 w e e Ue a an a axial ocations in the e, of the maz~ifue rod p dens (ku a given fuel a y and axial location to thc limiti fuel power densit kM/ft at that location p,PLH&iL 4~a a a-The is applicable to a specific planar height and is equal to thc sum of the for all thc fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.LH PRc PAGE AMENOMEMT HQ.M S g i</-~;~/~-c&K g.g/c....>'"iP~X<$'<S~4 gEa Pa~d~~/r~Pad gk~~g X ic PW g pQ f' 0 0' Qp<>A>4'cd')o~s~<)g4tr i+;a, gns 0(5 440 cllhlP4 Fo~5~gLpg 4L~Il ('t$wo+'.t.)e4t rtp<.'~-ka instr>ment functional teat means the icQcctlon of a simulated signal into the inatnmcnt primary sensor to verify the proper inst res e ala o inltiatl action S4 I CHANC l WA(4(NOR OSCCQe44 ko~lltative by observation of behavior during operation. This determination shall include, vhere possible, comparison of the eycndcnt instruments measuring the same~We (fy 4~cle~)),c a teat o relays and contacts of a logic circuit~Aq~~)~X'llxEC P I.5~@trip sys cm m ans an arrangcaLcnt of instruILcn channel trip sioxals and auzillary egiyaent required to initiate action to accomplish a protective trip function.trip aystea may require one o more instrument el trip signals related to one or more lant parameters order to tiate trip system action.I tiation of protec ve action require the tripping of a si e trip ayatca or co ident trlyying of tvo trip sy caa.70-kn action lnitiat by the protection system vhen a t ia reached.k protective a tion can be at a channel o system level.v<<k system yrotective action vhich results from the yrotec ve action of the channels monitoring a particular plant ndition.BFI Unit 2 1.0-8 AMENDMENT NO.2 y 7 PAGi=UF~~

, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.PAGE Al i 5 eCiCCa e f.l OCT 2 11993 9.S ated uto a ctua o-Simulated automatic actuation me applying a simu ted signal to the s or to act ate the circu t in question.A0<ement of relays, ontacts, other decision output.0.$~o A logic is an ar compon ts that produces (a)I iat-A logic hat receives signals from charm ls and roduceh decision utputs to the actuat on logic.P eNAfMeee Chl,f~oft)<~p p sreg C'4+c4f~ssepeC OC-A logic that receives signals (ei er from initiati logic or charm ls)and produces decis on ish a rotective action ,A:~gfiall be the ad)ustment, as necessary, of the channel output such that it responds necessary r e and accuracy to known values of the parameters-vhWh the~cf channel monitors.The@bann~ca bratio shall encompass th entire channe, nc u a a trip functio , shall may be erformecC by any series oP$equential, overlapping or total c~a~ne~~p.$sucn tnat tne entire channel is~Calibrated. Hon-ca rata e components s be exclud rl from his.rendu rement, bM will be included in~nel functiomg test and source check.jL CJ/A44~4 Rr~c l)ofeAc 11 begs.,a,B ,6 led p 4C di>i4 e niceties f a a latedgei into the channel as close to the sensor as practicable to verify OPERABILITY including ala 1~trip functio she.CHAdfogf F4spCnaqf4C. &rr~<k 4 W Q w csC~st~its aC~pfdip>o eÃe,pp~t e<~~C+fitf Se W4 C'~C~fd4e44.st e-t e e on o a s mu a e y n the sensor t verify OPE ITY includ ng alarm~and/or ri f ctions.~ce ec e e qualita e assemagnt of res nse when e channel or is expo ed to a radioact ve so or mu iple of sources.Adf pAGE~oF=BFH Unit 2 1.0-9 AMENDMENT NO.2 y'7

(gal (C-A functional st is the manual operation or initiation of a system, subsystem, r component to verify that it unctions within design tolerance (e.g., the manual start of a co spray pump to verify that it and that it pumps the requir volume of water).g~Qg~-e reactor is in a shutdown conditx when the reactor mode itch is in the shutdown mode posit n and no core alterations are ing performed. -An engineered safeguard is a safe system the actions of ich are essential to a safety aetio quired in response to cidents.-A report e event shall be any of-hose conditi specified in Sectio 50.73 to 10 CPR Part 50.&elate&BB.~$3 Shall contain the methodology and para ters used in the ca ation of offsite oses resulting from ra oactive gaseous iquid effluents, the calculation of gase and liquid ef flu t monitoring Ala/Trip Setpoints, and in he conduct of the vironmental Radio ogical Monitoring Progr'The ODCM shall al contain.(1)the Radioactive Effluent Con ls and Radiological Enviro al Monitoring Programs (2)descriptions o the information hat should be included the Annual Radiological Bnvironmental rating and Annual Radi ctiva Bffluent Release Reports requir by Specifications 6.9.1.and 6.9.1.8.CC.e controlled process of charging air or gas from the primary ontainment to maintain te rature, ressure, humidity, co entration, or other operat g condition x such a manner that re acement air or gas is requ ed to p ify the containment. DD.)fan~~-controlled process of d charging air or gas from the primary c tainment to maintain te rature, pressure, humidity, conc ration, or other operat condition in such a manner that repla ment air or gas is not ovided or required.Vent, used in syste names, does not imply a venting process.BPN Unit 2 1.0-10 ppG'E/8 0 0 I 1.&-Shall be that ine beyond which t ed, or otherwise con rolled by TVA.~~land is not owned, 1~l HH.IZ.-Any area at or yond the SITE BO Y to which ccess is not ntrolled by the lic ee for purposes of pr action individuals f exposure to zadiati and radioactive mate als or y azea within he SITE BOUNDARY used or industrial, coaeer l, ins tutional, or ze eational purposes.~jC IPOC,WI'tJ C%a+4-XhQ SE EQVZViLLENT Z-131 shall be the concentration of I-131 (4a~gm)a oae would produce the same thyroid dose as the quantity and isotopic mixture of Z-131, Z-132, I-133, Z-134, aad I-135 actually present.The thyroid dose conversion facto@used for this calculation shall be those listed ia Table IZZ of TZD 1(844 iCklculatioa of Distance Pactors for'ower and Test Reactor'ites JJ.~P.ltd-The char adsozber vess ls i tailed on the disc zge of the steaa get a r effector to y ide de y to a unit's offga activity prior to rel ARN.aay per defined individual ia a coatro led or URlSSTRI ver, aa indivi is not a MEMBER Ot%BRIC during ia which the indi idual receives an oc tional dose (as 10 CPR 20).-Surveillance R zements shall be met ing the OPERATICN. ITIOHS oz other tioas syecified for dividual limiting coadi ioas for oyeratioa ess othexQise stated aa individual S illaace Requirement .Each Surveillance Requirement shall be perfo within the specif ed surveillance interval ith a maximua allowable ension aot to ex 25 yercent of the sye fied surveillance int.It is not int that this (extension) provisica be used repeatedly as a conv ence ta extend surveillance tervals beyond that~specified for suzve llances that are not foaae4 during refue ing outagu.of a Suzveil ance Requireaeat wi the specified time ahall constitute compliance and Oy ILITY requizements for 4 ask condition for ration aad associa action statements t4M o se required these syecificati .Surveillance ~~I!it ia di red that a s eillance was not per ormed within."s specified fr cy, then comyl ce with the requiz t to decl~"e the LCO not met y be delayed, the time of dis very, up to 24 hours or uy to t limit of the s cified frequency, w chevez.s less.This delay iod is permitt to allow perfozmaace of:.".e surveillance. BPN Unit 2 1.0-11 0,, TY 373 Specs ji c14i~I J period, the LCO e condition(a) the ot Zg tbe surveillance is not pe formed within the del ately be declared n met, and the appli ace he ared.the aurve llance is performed ithin the delay.period s eillance is t met, the LCO taus immediately he declar met, ancL the appl able condition(s) t be entered.-Surve lance Require ta for Inaervice T ting of AQ%Code Clas 1, 2, 3 componeata ll be applicable aa ollcwe: 2.Inae ce testing of Code Class 1, 2, 3 pape aad valves hall be perfo ia accordaace with ection XZ of the AQC Boi er and Pressure ode and applicable aa required 10 CPR 50, Sec oa 50.55a(g), excep chere specific 149ittea re ef haa been gran ed hy the ComsLiaaioa. auaat'to 10 CPR 50, tion 50.55(g)(6 (i)~Surveillance tervala apecif ied Section XI of the Boiler aad Pres e Vessel Code applicable Addlmda r the iaservice test Lctivi'ties requir hy the AQQ Boiler ressure Vessel C and applicable shall be appli le follcws ia thea techaical apecifi iona'oiler aad Pre Vessel R f reymciea Code and ayplicahle'or oraiag inaervice te logy for iaae ce Y thly Quarterly or every 3 mmthl S~iaamaally every C mon Ivery 9 tha Yearly or ly At least yer 7 days At least per 31 days At least once 92.days At least caco 184 days At least once pe 276 days At least oace yer 66 days 3 0 ymviaioaa of.ificaticcL 1.0.are applicable to he required frequ iea for perfo iaaervice teatin acti ea.perf of the above ervice teat activities shall he ia additi to other specif aurveillaac requirements. 5.thing in ASMS Boiler aad assure Veaae Code shall be trued to s rsede the r emeata of any echnical s ifica'tice BPS Unit 2 6.The ervice ction prograa fo piping identiti d in NRC Generic tter 88-0 shall he perfo in accordance ith=.".e staff yoa tiona on s e, methods, raonnel, and sample expansion luded in a generic letter.I 1.0-12 pAc;~~IOF~ , A1 Sa e u do NOV p g)ggg FS has developed Appends R Sa Shutdown Prog aa.Thi ogram is to ensure that the equipm t required by he hpp R fe Shutdown-Analy s is maintaine demonstrat func onal as follovs: 5z~1.The unctional requirem ts of the Safe utdown syst and~equi t, aa veil aa app opriate compensa ory measures should ese systems/compon ts be unable to erfozm thei ntended ction are outlin in Section III f the Progr (@so 2.Tes and m toring of the hp endix R Safe Shu ovn syst and equ pment are defined Section V of Prograa.*-/'-The COLR is the unit-specific document that provides for the current cycle.These cycl~pecific ~re o cq W g-s shall be determined for each operating cycle in accordance vith Specif icatio~~Plant operation within these limits is addressed in ndividual specifications..C.S-h 1 ting control patte~shall be attern vhi results in ore being.on the~t f i.e.crating on a 1 isLiting value or hPLHQR f f or NC R.P(3 A pngu)2NfE'R7 Z-C/pe~)BPS Unit 2 1.0-12a IMENOMENf SL 2 4 I PPQ I 0> 0 INSERT 1'Page 0 of 3)ACTIONS LEAKAGE Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated conditions within specified Completion Times.LEAKAGE shall be: a.Identified LEAKAGE LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank;or 2.LEAKAGE into the drywell 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;b.C.d.Unidentified LEAKAGE All Leakage into the drywell that is not identified LEAKAGE;Total LEAKAGE Sum of the identified and unidentified LEAKAGE;Pressure Boundar LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS)component body, pipe wall, or vessel wall.The definitions found in this insert will be placed in alphabetical order with the other ISTS definitions. PAGE

INSERTS (Page 2 of 3)LINEAR HEAT GENERATION RATE (LHGR)MODE The LHGR shall be the heat generation rate per unit length of fuel rod.It is the integral of the heat flux over the heat transfer area associated with the unit length.A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table l.1-1 with fuel in the reactor vessel.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: SHUTDOWN MARGIN (SDM)a.Described in Chapter 13.10, Refueling Test Program, of the FSAR;b.Authorized under the provisions of 10 CFR 50.59;or c.Otherwise approved by the Nuclear Regulatory Commission. SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: 'a~b.C.The reactor is xenon free;The moderator temperature is 68'F;and All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

INSERT 1 (Page 3 of 3)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 TURBINE BYPASS SYSTEM RESPONSE TIME THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components: 'a~b.The time from initial movement of the main turbine stop valve or control valve until 80%of the turbine bypass capacity is established; and The time from initial movement of the main turbine stop'valve or control valve until initial movement of the turbine bypass valve.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. 'I'I INSERT 2 Defin'itions

1.1 Table

1.1-1 (page 1 of 1)MODES MODE TITLE REACTOR MODE SMITCH POSITION AVERAGE REACTOR COOLANT TEMPERATURE ('F)Power Operation Startup Hot Shutdown()Cold Shutdown()Refueling(b) Run Refuel(a)or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel NA>212<212 NA t (a)All reactor vessel head closure bolts fully tensioned.(b)One or more reactor vessel head closure bolts less than fully tensioned. BFN-UNIT 2 1.1-7 Amendment pAGE~OF IMSERT 3 (Page 1 of 21)Logical Connectors 1.2 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~N and Qg.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 indentions of the logical connectors. Mhen 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. EXAMPLE S The following examples illustrate the use of logical connectors.(continued) BFN-UNIT 2 1.2-1 Amendment PAGE~OF~ 0 0 INSERT 3 (Page 2 of 21)p(f)Logical Connectors

1.2 EXAHPLES

(continued) ACTIONS CONDITION REQUIRED ACTION COHPLET ION TINE A.LCO not met.A.l Verify...A.2 Restore...In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.l and A.2 must be completed. I (continued) BFN-UNIT 2 1.2-2 Amendment INSERT 3 (Pago 3 of 21)Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES ELNN.(continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.LCO not met.A.l Trip...A.2.1 Verify...A.2.2.1 Reduce...+0 A.2.2.2 Perform... A.3 Align...This example represents a more complicated use of logical connectors. Required Actions A.l, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector gg 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 gg indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed. BFN-UNIT 2 1.2-3 Amendment PAGE~+OF~ INSERT 3 (Page 4 of 21)Completion Times 1.3 1.0 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 Times(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 HODE 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.Nhen 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 divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will~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.(continued) BFN-UNIT 2 1.3-1 Amendment PAGE~oF~>>

INSERT 3 (Page 5 of 21 p(<Completion Times

1.3 DESCRIPTION

(continued) H, 1 RlhhtdHI, byt, p t, 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 inoperabi1ity: a.Must exist concurrent with the~fi~t inoperability; and 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 division, 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 mqdified"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..." Example 1.3-3 illustrates one use of this type of Completion Time.The 10 day Completion Time specified for Condition A and 8 in Example 1.3-3 may not be extended.BFN-UNIT 2 1.3-2 (continued) Amendment OF 0 0 INSERT 3 (Page 6 ef 21)pn'.3 Completion Times{continued) Completion Times 1.3 EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME B.Required Action and associated Completion Time not met.B.l Be in HODE 3.ANIN B.2 Be in HODE 4.12 hours 36 hours Condition 8 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 HODE 3 within 12 hours~N in HODE 4 within 36 hours.A total of 12 hours is allowed for reaching HODE 3 and a total of 36 hours (not 48 hours)is allowed for reaching HODE 4 from the time that Condition B was entered.If HODE 3 is reached within 6 hours, the time allowed for reaching MODE 4 is the next 30 hours because the total time allowed for reaching MODE 4 is 36 hours.If Condition B is entered while in MODE 3, the time allowed for reaching HODE 4 is the next 36 hours.(continued) BFN-UNIT 2 1.3-3 Amendment PAGE~OF~~ lNSERT 3 (Page 7 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETIQN TIME A.One pump inoperable. A.l Restore pump to OPERABLE status.7 days B.Required Action and associated Completion Time not met.B.l Be in MODE 3.~AN B.2 Be in MODE 4.12 hours 36 hours Mhen 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 8 is entered, Condition A and B are exited, and therefore, the Required Actions of Condition B may be terminated. Mhen 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.awhile 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.(continued) BFN-UNIT 2 1.3-4 Amendment PAGp pg OF~ 0 INSERT 3 (Page 8 of 21)t}.(Completion Times 1.3 EXAMPLES DNN3.R{ti d)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.(continued) BFN-UNIT 2 1.3-5 Amendment PAGE~OF~

INSERT 3 (Page 9 of 21)Completion Times 1.3 EXAMPLES (continued) AMP.3-3 ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME A.One Function X subsystem inoperable. A.1 Restor e Function X subsystem to OPERABLE status.7 days~AN 10 days from discovery of failure to meet'he LCO B.One Function Y subsystem inoperable. B.1 Restore Function Y subsystem to OPERABLE status.72 hours 10 days from discovery of failure to meet the LCO C.One Function X subsystem inoperable. One Function Y subsystem inoperable. C.l Restore Function X subsystem to OPERABLE status.+0 C.2 Restore Function Y subsystem to OPERABLE status.12 hours 12 hours'I (continued) BFN-UNIT 2 1.3-6 Amendment pAGp BF Ot'~~ 0 INSERT 3 (Pago 10 ot 21)Completion Times 1.3 EXAMPLES BNNP.3-3 t tl 4)When one Function X subsystem and one Function Y subsystem are inoperable, Condition A and Condition B are concurrently .applicable. The Completion Times for Condition A and Condition B are tracked separately for each subsystem, starting from the time each subsystem was declared inoperable and the Condition was entered.A separate Completion Time is established for Condition C and tracked from the time the second subsystem was declared inoperable (i.e., the time the situation described in Condition C was discovered). 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.l 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 subsystem was declared inoperable (i.e., initial entry into Condition A).The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met.In this example, without the separate Completion Time, it would be 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.The separate Completion Time modified by the phrase"from discovery of failure to meet the LCO" is designed to.prevent indefinite continued operation while not meeting the LCO.This Completion Time allows for an exception to the normal"time zero" for beginning the Completion Time"clock".In this instance, the Completion Time"time zero" is specified as cowaencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.(continued) BFN-UNIT 2 1.3-7 Amendment

t 0 1.3 Completion Times lNSERT 3{Page 11 ef 21)Completion Times 1.3 EXAMPLES (continued) AHP.3-4 ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME A.One or more valves inoperable. A.l Restore valve(s)4 hours to OPERABLE status.B.Required Action and associated Completion Time not met.B.1 Be in HODE 3.B.2 Be in HODE 4.12 hours 36 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 fs 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.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 (plus the extensions) expires while one or more valves are still inoperable, Condition B is entered.(continued) BFN-UNIT 2 1.3-8 PAGE

INSERT 3 (Page 12 of 21)e)Completion Times'.3 EXAMPLE (continued) ACTIONS NOTE"Separate Condition entry is allowed for each inoperable valve.CONDITION REQUIRED ACTION'OHPLETION TIME A.One or more A.l Restore valve to 4 hours valves OPERABLE status.inoperable. B.Required Action and associated Completion Time not met.8.1 Be in NODE 3.8.2 Be in NODE 4.12 hours 36 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.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.(continued) BFN-UNIT 2 1.3-9 Amendment I~~~~~~~I I~~I~'I~~~~~~~~~~~I~I~~~~~~~~'~~~~~~~~~~~~~~I~'~~~~~I'~~I~~I~~~I I I I'tl'~~I I s~~I I I.I~~~

INSERT 3 (Page 14 of 21)Completion Times 1.3 EXAMPLES ttttt.tt I tt tt 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.l 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.l or A.2 is met, Condition B is exited and operation may then continue in Condition A.(continued) BFN-ONIT 2 1.3-ll Amendment pAGE~~OF~ 0 INSERT 3{Page 15 of 21)Completion Times 1.3 EXAMPLES (continued) ACTIONS-7 CONDITION REQUIRED ACTION COMPLETION TIME A.One subsystem inoperable. A.l Verify affected subsystem isolated.1 hour Once per 8 hours thereafter A.2 Restore subsystem to OPERABLE status.72 hours B.Required Action and associated Completion Time not met.B.l Be in MODE 3.B.2 Be in MODE 4.12 hours 36 hours Required Action A.l 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.l.If after Condition A is entered, Required Action A.l 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.l (continued) BFN-UNIT 2 1.3-12 Amendmen PAGE~OF

INSERT 3{Page 16 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES BNIELLl.3.7 ('i dl 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 Ae2 has not expired.IMMEDIATE When"Imaediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.BFN-UNIT 2 1.3-13 Amendmen wo,l:a E5<~ t 1.0 USE AND APPLICATION

1.4 Frequency

IMSERT 3 (Page 17 of 21)Frequency 1.4 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 Limiting Condition for Operation (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, 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.Example 1.4-4 discusses these special situations. 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 (continued) BFN-UNIT 2 1.4-1 Amendment INSERT 3 (Pago 18 of 21)Frequency 1.4 1.4 Frequency DESCRIPTION (continued) criteria.SR 3.0.4 restrictions would not apply if both the following conditions are satisfied: a.The Surveillance is not required to be performed; and b.The Surveillance is not require'd to be met or, even if required to be met, is not known to be failed.EXAHPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown)is HODES 1, 2, and 3.,4-SURVEILLANCE REQUIREHENTS SURVEILLANCE FRE(UENCY'erform 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 interval specified in the 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 HODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not (continued) BFN-UNIT 2 1.4-2 Amendment lllSERT 3{Page 19 of 21)Frequency 1.4 1.4 Frequency EXAMPLES~4.)ii d)otherwise modified (refer to Examples 1.4-3 and 1.4-4), 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 NODE or othe'pecified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the HODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.Np 4-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits.Once within 12 hours after>25%RTP 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"~AN" indicates that both Frequency requirements must be met.Each time reactor power is increased from a power level (25K, 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 extension allowed by SR 3.0.2.(continued) BFN-UNIT 2 1.4-3 Amendment PAGE~a"~ INSERT 3{Page 20 of 21)Frequency 1.4 1.4 Frequency EXAMPLES BNN4-2 I tt dl"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 25K RTP.4 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. A th tt diff hl tl d~f lth 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>25K RTP to perform the Surveillance. The Surveillance is still considered to be within the"specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by 3R 3.0.2)interval, but operation was<25K 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 NODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power>25K RTP.(continued) BFN-UNIT 2 1.4-4 PAGE Amendment 0 0 INSERT 3 (Page 21 of 21)Frequency 1.4 1.4 Frequency EXAMPLES~4.t ti d)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, there would then be a failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.X.4-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-NOTE Only required to be met in MODE l.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 (plus the extension allowed by SR 3.0.2)interval, 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.BFN-UNIT 2 1.4-5 Amendment S 0 THIS PAGE I ZOHALLY LE BLAHK BHf Unit 2 1.0-12b AMENDMENT NO.1 9 2 PAGED=~

~)<ciP~ctlion /.I MAY fg]88g hift)At east once er 12 ho s.(Da)At lea t once pe normal len r 24 hour ay (midni t to mi ght W (WeelQy)M (thly)Q (guar rly)Sh (Semi-ally)T (arly)(Ref ling)S/U (Start S.A.P (Pri)At least on per 7 d s.t least once r 31 day At east once per months o 92 days.At le t once per 6 m ths or 1 4 days.%east ce per year o 366 days.At ast onc er operating ycle.Prior t each re tor startup.ot applic le.Co pleted prior to each lease.BFK Unit 2 1.0-13 AMBfDIWIIENT go g8 5

THIS PAGE INT ZONALLY LEFT B BFN Unit 2 1.0-13a AMENDMENT NO.)6 PACi~t-

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Al.0 vs e 4Phscevwd Afore McLcgncd~w<<, speeiA+y n I.I SCc+on 4f I n;hlkhg fi'Pn A 2.p4R aieiip eu iimk PCS'S h'Nt ,'g'c n S ts ar limits elov ch'h r onable iat e of th claddi and pr ry sy terna e as ed.Ex ediag s ch a 1 t requi s unit utdovn and r iev by e Nucle Regulat Commi ion be re res tion o unit oper ion.Op ation b ond such a limit y'not itsel resu t in se ous consc caces b it ind ates operati aal def ien sub)ect to regula ry rcvi B.-The limiting safe y system s ti1lg are ettings instr'atioa ch ini ate thc utomatic O pro ective ac ioa at a evel su that th safety 1 ts vil not bc exes dcd.The egion be een the afety 1 t and th e sett s reprc ts marg vith no opera on lying elov the setti s.Thc mar has be establis so vith pr er opera oa of thc instrumentation the fety lim ta vill ever be eedcd.C.-The limit conditions or opera on spec the acceptab levels of stem p formaace ecessa o assure s e startup opcratio of the fac ity.Mh these c itions are et, the ant can be eratcd safe aad abno 1 aituat can be fely coa oiled.In the ev t a Limitiag C itioa and/or sociated reguircaen cannot be sati ied because of ircumstanccs in excess of oae addressed the apecifica on, the unit shall be p aced ia at less Hot Standby vi hours and in Co d Shutdova vith the folloving 30 hours ess corrective asures are comp-tcd that pc ope tioa under the rmisaible disco ry or until react r is placed in operational co Cion in vhich e specif ation is not app cable.Except ns to these rcquir ts shall be stat in the indivi al specifics oas.This provi action to be cn for rcumstan s not directly pro ided for in thc sp ificatio and vhere occurr ce vould viola the int of the pecification. For example, if a specif ation lls for tvo syst (or subsystems to be operable pro ides for explicit r uirements if o system (or ubsys)is inoperable, t if both syst (or subsyst)are noperablc the unit s to be in at least Hot St in hours and in Cold utdova vithin the folloviag 3 ours if the inoperable co ition is not corrected BFH Unit 3 1.0-1 PAGE W OP'/~ ill SPeC,q'myon I t NOV is la 24 en a sf stems s spstcmy train~components or deaf ce is d ermfned to be operable solely because Lts onsft pover s ce is Lnoyerabl, or solely because its offsite ver source fs inoperable, it ma be considered era e for the rpose of sat s the requfr ent of Lts ay licab Lfmftiag ondition For Oyera on, provide: (1)i s corr sponding o fsite r diesel pover s urea Ls oyerable;and (2)all o its re undant stem(s)aubayst (s), train(s), co ent(s), devi e(s)ar operab , or 1Dc ae sitfsty these r qufrement.Unle s both onditi (1)and)are satisfied, e unit s 11 be y aced Ln at lees Hot Stand vi 6 hours, in at east Col Shutdo vi the fol ovtna 30 hour."This d finiti is not pylicab e in Cold Shut or Refuel.This provfsf descri es vhat additional condi ons must be atisfied o perai oyerati to c tfnue consis t vith the yecifica ons for er so ces, en an'offsite r onsite r source s not o rable.It syec ffcally yrohibits operation vh one df sion is perab baca its offsite or diesel pove source f inopera~and a stem, sub~stemf rain f comyon t f or de Lce fn ther d ision yerable f r another re on.Thi provis yermft the r irements sociated vi indfvf a@st, subay tens, tr f coayon ts, or devi es to be onsisten vith th r resents of the associat electri al yover source.t all oyeration to be govern by the fme lfmf of the'requf ts asso ated vith Lfmf t Conditi For~Oyerat for the ffsite or di'e el yover ource, not the fndMd requfrem ts for each stem, subsystem, train, coayonen or device t is determined to be inoperable solely because o the inoyer flit@of fts offsfte or diesel yover source-Pr to a the firs control rod r n PhC~iriSea A sp's em'u sfstemy component, or d~ce shall be qigmhli or e o erabf lite vhen it is capable of pitfOEILtSg ftS IpOCOOflNL fORC SOS(~~1 n ss attendant fnstamentatf on controls, no emerency electrical ove~so~cool%ay er seal vater, lubrication,& other amilfary ay6paeat that are required for the system, subsystem, cmyoeent, or device to perform its unction(s) are also c able of yerLoraiag their related support function s.d in'~F.ft rat me that a stem or functio Ln i requir er en s p fo ng g3o c.-Iamc te qu red a tion vi 1 be tiated soon practic le consid hg the afe op atio of unit the e of ha re ui aetio BFS Unit 3 1.0-2 NIENDMENT NO pr,GE~oF

~(gi~H.on NOV TS 1988 V eactor y er operation is aalu op ration/HOT ST Y or RUH NO vith the r ctor crit al and a ove 1 per t rated po r.-The rea or is in the TARTUP CO ITIOH vh the wi raval of c trol rods fo the purpose f making t e reactor crit al has beg , reactor po r ia less oz equal to 1 percent of ra ed and the e sin eS BY DE.-The reactor is in the HOT STAHDSY COHDITIOH vhen reactor over is less or equal to 1 p cent of rated.The eactor is in e SThlKUP/HOT kHDBT MODS, and e reactor is not the STARTUP HDITIOH.The r ctor coolant tern rature may be gr ter than 212 t.Hote hat a HOT ST BY COHDITIOK c t exist simultan ously vith a SThR COHDITIOH due to the difference intent.h HO STANDBY COHDITI exists vhen e reactor mode s tch is placed in the SThRTUP/SThSDBY posi on{for example, o comply vith LCO)and yover evel has been duced to 1 percen or lover.Ament e control rod are being vith avn for the yurp e of increasing reactor pove level, the reac r mode svitch been placed in the SThRTQP/HOT STAHDBY position, and reactor pover level is at or belov one percent, a ST COHDITIOH exists.res+or ik th SHUTDOWNS HDIT 8 vhen t e rea or is in a Sh dovn o R ue o W2-The reactor is in the HOT SHUTDOMR C ITIOH vhen eactor coolan temperature is reater than 21 F and the actor is in th SHUTDOWNS COHD 105.-'The react is in the C SHUTDOWH COHDITI vhen react coolant tempera re is equal o or less 212 t and the reactor is tn the SHUTDON COHDI 05.The react s in the LD COHDI IOH vhengeactor coo t temperat e ia equ to or less 212 in aap+de of Oper on{except~defined K.2 above).BFH Unit 3 1.0-3 AMENDMENT i[5.y p.9>>'g~>>~~C>>4 M.-The reactor mode e<<itch of Operati of the actor<<hen t re veaae , except tha the Mo of Operation the re ctor mode e<<tch ia t porarily mov permitt d by the not.When here is no f vessel, he reactor i conaide d not to be or operat onal conditi.The r actor mode oeiti or ma be inoperable. APR 3 positi determines the ia fue in the r actor y rema unchenge<<hen to anoth r poaiti as 1 in the actor any Mode f Oper tion e<<ch may then be in l 20 3~-The reac r is in the ThRTUP/HOT S BY MODE~~en the rea r mode s<<itch ia i the"SThRTUP/8 SThN5BT" pos1txcni. a xa of ten ref red to aa just e SThRTUP MODE.-The reactor in the Run Mode<<h the reactor de s<<ch ia in the"R" position.-The reactor i in the Shutdo<<n Mode<<de e<<itch ia in the"Shutdo<<n" positi.hhalhh-reactor mode a<<itch ia in the"Refuel" position Sce XvSHkcM'o~ 5 r~Scdw4 fsys s~~Q,g The reactor mode s<<itch may'e placed in any position to per orm required testa or maintenance authorised by the shift operations supervisor, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.~The reactor aode s<<itch may be placed in the"Refuel" position<<hile a single control rod drive ia being removed fry the reactor pressure vessel per Specification 3.IO.h.5 provided that reactor coolant temperature ia equal to or less than 212~F.The reactor mode s<<itch may be placed in the"Refuel" position<<bile a single control rod is being recoupled or<<ithdra<<n provided than tbe one-rod-out interlock is OPERABLE The reactor mode s<<itch may be placed in the"Startup/Hot Standby" position end<<ithdra<<al of selected control rods ia permitted for the purpose of determining the OPERhBILITY of the RWM prior to<<ithdra<<al of control rode for the purpose of bringing the a or to criticalit ~~~~+'<~<4 Ch s BOA 15',3~~BFH Qnjt 3 1.0-4 mENDMarr aO.I 6 9 pAGE 5

RTf'~g Q q~E'ER cucr yea+r<c imam+'on I, I nSR<raw W~T<cKttr Cogafft of 3gQ3 Nwt.NAY 2 D 1993 7HCRp v-Rated pover re crs to opcratio at a reactor over of 293 fata th is a 1 so tcrme 00 pcrccnt pov and is the aximum po r level aut rized by thc o rating license.Rated steam lov, rate coolant flo rated neutron ux, and rated n lear cyst prcssu refer to the clues of these arameters vhea the reactor s at rated er.-Primary coatainm t integrity means that..the d ll and pressure sup cssion chamber a e intact and all f the follovi conditions are sat fied: 1.hll nonautomat c containment isolat n valves on lin connected o the reactor olant system or cont ament vhich ar not r uircd to be op during accident co itioas arc clos d, except fo valves that ar opea under administr ive control as perm tted by Spccifi tion 3.7.D.ht less one door in ea airlock is closed sealed.1 automa ic contaiamcnt is ation valves are OPERhBLE or each, li vhich ntaias an iaopera le isolation valve is isolated as equ cd by ecification 3.7.D.2.aa all lf f land and nanaaya aca clcaad.Secondary containm integrity means that e required unit eactor zones aad rcf iag zone are intact the folloving c itions are met: a)h least oae door ia ea access opening to the urbine bui agi control bay aad ut-of-doors is closed.b)The st gas treatmeat sys is OPERhSLE and can intain 0.2S inche of vater negative p ssure in those areas v rc econdary co ainmcnt iategrity i stated to.,exist. P~c)hll econdary co aiamcnt pcnetrati required to bc closed duri accident co itions are either: 1.Caps e of being c ed by an OPERASLE condary contai cnt automatic solatioa position, r 2.Closed by least onc se ndary containment tomatic isolation va e deactivated n the isolated pos tion.2.Reactor one secondary c tainmcnt inte ity means the unit reactor bu ding is intact d the follov conditions are met: a)ht least on door bctveen opening to t e turbine building, control bay and out f-doors is closed.BFH Unit 3 1.0-5 AMENDMERT HO.I 7 0 pAGE~OF~ 0 tlBXI4tk (P.(Cont'd)Sfec'~1.1 NOV 18 t888 2.b)The tandby Gas Trea cnt System is 0 RABLE d can maintain 0.25 ches vater nega ve pressure on c unit zone.c)hll the it reactor buil ng ventilation stem ctrations required t be closed duri accident condit ons a either: l.Capable being closed b an OPERABLE rea nr bu lding ventilatio system automati isolation syst , or 2.Closed by at ast one reactor uilding vcntil tion stem automat isolation valve deactivated in the i lated positio If it is des able for oper ional considers ons, a react r zone may be isolate from the othe reactor zones d the refuel zone by maintaining least one cl cd door in each ommon passagevay tveen zones.*eactor zone sa ety-related fca res are not co romised by op ngs betvecn a acent units or efuel zone, unl s it is desire to isolate a g ven zone.3~Refue xone secondary c tainment inte rity.means thc refuel zone is inta t and the follov conditions e met: a)ht lc t one door in ca access openi to the out-of-doors is clos d b)The stand gas treatment tea is 0 and can maintain 0.25 inches vater negative p ssure on the fuel zone.c)hll refuel x ventilation sys em penetratio required to e closed duri accident condit ons are either 1.Capable of be closed by an refuel zone ventilation sy ea automatic is ation system, or 2.C sed by at less one refuel zon ventilation system aut tic isolati valve deactiva d in the isolated posi on~Xf it is desira le for operati 1 considerati ns, thc refuel one may be isol ed from the r ctor zones by aintaining all ches in place b tvecn the refu 1 floor and reactor zones and at cast one closed door in each a css betveen the refuel zone and e reactor buil ng.*Refuel ne safety-related features are no compromised b openings betve the reactor building unless t is desired to isolate a giv xone.*To effect vely control one isolation, 11 accesscs to the affected zone vill be locked o guarded to revcnt uncontro ed passage to the unaffected zones.PPidF~<i;,.~ BFE Unit 3 1.0-6 NENDMENT NL', P g 0 ~p,i 1.8'ggi ligfe~~OM NY208%0 d-Interval~etveen tha end of ene refueling,outage'or a p itular unit and the end of the next esihaequent refueling utagc for he same un S./o(L i~)g hg cd~M i0M fgu'-TNZWrel m;u-Refueling outage is the period of time etve the shutdown of the ua prior to a refueling and t startup of th uni after that refueliag. For thc purpose of dcsigna frequency of tes ng and survcillancc, rcfucliag outage shall mc regular scheduled outage;hove r, vherc such outages occu thin 8 moaths o e completion of thc previous refueling outage, the required surveillance testing need not be performed until the next c ularly scheduled outage fOW<+6n C ately(i-CORE ALTERATIOH shall be the movement of any fuel, sources, reactivity control components, the reactor vcsscl vith the vessel head removed and fuel in the vessel Movement of source range monitors, iatcrmediatc range monitors, traversing in-core probes, or special movable detectors (including uadcrvessel replacement); 4a-ace~.0 Suspension of CORE ALTERATIOHS shall not preclude completion of movement of a component to a safe e P4C;4'g~g f t~,pn'~~o~ 4 jQ>>a Qlf 0/6 Q 4tf f Od'ega C4/<<O+e s e u-ess othervisc in cate, rotor vessel press s listed in thc Techn pccificatioas arc those measured e rea or vessel steam space detectors ,((lM W S~~(l<<fI 1 Cc(w)F4(-i~>~e'<g~cor C~~Rf'Seh.is calculated to cause some point in the assembly to experience boil transitioa, M thc actual assembly operat pover.ll g ogpu ghe 4oaa O P opfnqr'ke chfr f.Lg 44)on o liag means e rcg e betv nucleate anil film boiling.Transition b+ling is the regime ch both nucleate aad fi&~iliag occam termittcntl vith ne be c letel stable.BPS Unit 3 34 ghcat, at o, or all fuel as es a a ocations in the co of thc maxiinua fuel od pover density/ft~~a given fuel ly and axial locat to the limiti el od pover density (t)at that location APL HGR, SfAk~44 applicable to a specific a AMENDMENT NO.I W HFc.p L,fff f~~4$4$w~ef/~~ghga eg iwv e~g ppierfr@lel Q/>>$4 A'."Q re~~A~-~>S 4>4 4 LHC4 CIfV(y<<(-4 g,'~f>>(<<g~3~~t~ri~(~~is</'(o CW>~m Z~O al at the specified height divided by the number of fuel rods in the fuel bundle.PAGE 70 o.'0 5 g.c!lic4AO I!I FEB 3 4m'~-CORE MAXIMUM I FRA IOH OF CRITICAL POMER the maximum value f the ratio of the fl-corrected CPR operat limit found in CORE OPERATING MITS REPORT divided by the actual CPR for all fuel ssemblies in the core.V.-An instrument calibration means the 5ustmcnt of an instrument signal output ao tha't it c responds, vithin acceptable rangey and accuracyf to a)ELovn val (s)of the parameter ch the instrument tora.-A channel is an arr ement of the acnso s)and aasociat components used to cv uatc plant variable and produce di crete outputs used in 1 ic.A channel term nates and loses it identity vhcre individ 1 channel outputs are combined in 1 ic.4~C~reN r+J o~4 s~s~aF4er i~A'ca4<z Or SSg!/e S dprj ltd-kn instrument functional test means the i+ection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action]/W M C IAWNNC!!by observation of behavior during operation. This determination shall include, vhere possible, comparison of thc cpendcnt instrumen s measuring thc sam rome cC f~a)lg Comiye&g/eg 4il rC-k 1~sic sysr f~gLto~t!!t a test of okg relays ang contacts of a logic circuit ao p,'I Ahar~~~y KPÃcT o'p+~y 5o!I d S l.o-g jh log(C, dr r~fs,eh.)oRlslI-o trumcnt channel trip signals and auxiliary equipmcnt required to tiate action to accomplish a protective trip function.A tr system may require one or morc instrument el trip ai la related to one or more lant parameters in rdcr to initi e trip system action.I tiation of protecti action may re ire the tripping of a s e trip system or th coincid t tripping of tvo trip ays 7'An action initiate by the protection stem vhen a limit a reached.A protective a ion can be at a channel or ays level.8~-A system protective action vhich results from the protective action of the channels monitoring a icular lant conditi BFH Unit 3 NENNENr N.gp~pgGF 0 0' , from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEH FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested. 0 0 Qi,'t ec.cc+on I FEB 8 4 t995 9.S u ated Automatic tuatio-Simulated automatic actuation me applying a simulated ignal to the sensor to actuate the ci uit in ques ion.10.o-A logic an arrangem t of relays, contacts, d othe components-t produces decision ou put.(a)logic that r ceives si ls from ch els d produces d ision output to the act ation logic P CgeegaL-CWL~gaawoo~ Ae req'ifcR RfiCDf (b)-A lo ic that recei es signals (either from init ation logic.or channels)and produces decision to accom lish a protective action..W M4 n hall be the adJustment, as necessary, of e uch that it responds~necessa~r e and accuracy to known values of the parameters ~h the~t*a lsseILI compass.the e e c ud ng a a an~tri unctio s all nclude t e i n tang.The ehanne,'i ration e er ones by any series of sequential, overlapping or total channel gg+~ygpg such that the entire channel is calibrated on~a recta Le components-salt~M-excluded gr~m s requirement, out wiil be incguded ini~hannhl functional teat aud.source che>L-8~~~PL&ruc 7~a<Te5 7 Dr Cdagal-Shall be: Rendu'TCd-the inJection of simu ted signal.,into the channel as close to the s or as practicable to verify OPERABILITY including alarm a~os-andCaeggcL4'lufcKefr. Ne<fAW~aH~Al 7kgr/4fkc g, p gttgnnenqDip >'Cm;j oC S~ert5a4gnfftlatfin9f~ WhSC-Cgygpfa.~y~ p~b.Bistab e Chanue-the i ection os simulated+ignal into the sensor t verify RABILITY+ncludi+ alarm/or tr functio 13.(Deleted)PAGE~3 BFR Unit 3 l.p-q i NENDMNT ND.I 90

@pi-A functional t t is the m nual operation or'nitiation of a system, subsystem, r componen to verify that t functions wit'n design tolerance (e.g., the manual start o a core spray pu to verify that it s and th t it pumps the equired volume f water).X.~AS-The reactor's in a shutdown con'tion when t reactor ode switch is i the shutdown mode p sition and n core alte tions are bein performed. -An en'neered safeguard is~q safety system the ac'ons of which ar essential to a safelry action required in res onse to accident Z.A reportable eve t shall be any of t ose conditions specific in section 50.7 to 10 CPR Part 50.BB.~(Q(s-S ll contain the me hodology and parameter used in the calcu tion of offsite dos s resulting from radio tive gaseous and'quid effluents, in t calculation of gaseo and liquid efflue t monitoring Alarm rip Setpoints, and in he conduct of the Vironmental Radiol'cal Monitoring Progra.The ODCM shall a so contain (1)the dioactive Effluent Co trois and Radiologi al Environmen al Monitoring Program required by Sectio 6.8.4 and (2)descrip'ons of the informatio that should be inc uded in the Annual Ra iological Environment l Operating and Ann l Radioactive Ef luent Release Reports equired by Specifi tions 6.9~1~5 and 6'..8.CC.-The controlled proces of discharging air r as from the prima containment to maint'n temperature, p essure, humidity, ncentration, or othe operating condition in such a manner that eplacement air or gas is required to pur'fy the containment. DD.EE.Yg~g-T e controlled proces of discharging air or as from the primary ontainment to maine'n temperature, pressu humidity, con ntration, or other crating condition in uch a manner that re acement air or gas s not provided or requx ed.Vent, used in sy tern names, does not'mply a venting process BFN Unit 3 1.0-10 pp@p Q OF~ t 1.V Q)Li Shall be t line beyond or otherwise ontrolled by TVA.the LancL ia t XZ.-Any area at r beyond the SZ BOUMARY to whi a ess is not co trolled by the li ensee for puzpos of protection of viduals fz exposure to radi ion and radioac ve materials r an area within e SITE BOUNDARY for industrie coaeercial. i titu ional, or rec ational purposes.~ic~aarig/ w~SS SQQIVALRNT X-13 shall be the concentration of I-131 Ckfya)alone would produce the same thyroid dose as the quantity and isotopic mature of I-131.I-133, I-133, I-134, and I-135 actually present.The thyroid dose conversion facto@used for this calculation shall be those listed fa Table IXI of TZD-14844 Calculation oC Distance tactors for Pcwer and Teat Reactor Sites.-The charcoal adsorbe>>tailed on the dis e of the steaa)et air e)ector de y to a unit's oC as activity prior to lease.ARR1.any peri ined r, an indiv in which the 10 CPR 20)~individual in a ccrc is not a ICl$%vidual receives an lied or CTRL TSRÃNLZC cupatioaal do (as-Surveillance R reeeats shall be t during the ORNA ZOSQI ZTZCN or othe conditices specifi for individual limit condi ions for operati unless othexiise a ted in an indivi S llance Requir ts.Each Surveill Requirement shall be fo within the fied surveillance erval with a RRX%1%oeable ension not to 35 percent oC specified surveill int o It is not ended that this.(ion)provisicÃL used tedly as a ence to extend'llance that if ied for illances., that are ot perfoaaect dur refuel outage.Qerfoxmnco of Surveill ce Requireaen within the specified time sball titute liance and ZIZTY requiremen s for Iialti339 ccodit for ope ation and as iated action stateme ts 0 tbearise r red by hose speci f i iona.Surveillance ts do not ve to be rformed on rable equipmen.I!it s discovered t a survei ance was no performed within xe,s specifi frequency, t complian with the r rement to decl are the LCO t met may be layed, free time oC discovery, up to 24 hours or to the limit f the speci frequen, whichever.s less.This delay period i permitted to aller per ormance of the surveillance. BPN Unit 3 1.0-11 PAGE ia OF 0 Zf the surveil ce is not performed withia the delay riod, the LCO icmsediatel be declared not m, aad the applicab coacU.tion(s) be catered.flgZ.whea the surveillanc is performed wit the delay period the surve laace is aot m, the LCO must i@lately be declar not met, an the apylicable condition (s)must entered.QI.Surveill e Requirements or ASMI Section X Pump and Valve Pr am-Surveill ce Requirements or Inservice Test of ASME Code Cla 1, 2, and 3 nents shall applicable as f lowsc Zaservice eating of ASMR e Class 1, 2, 3 pumps and valves shaE be performed in ccordance with S tioa XZ of the ASS Boiler Pressure Vesse Code and applic e Addenda as required by 1 CPR 50, Section.55a(g), except re specific written relief s beea granted the Coaad.ssion suant to 10 CFR 50, Secti a 50~55a(g)(6)(i).2~S eillaace inte als syecified in Se tice XZ of the Bo er and Pressure Vessel Code and apy cable AddexuIa f inse ice testing ac vities required by ASMI Boiler Pres e Vessel Code applicable Addenda shall be apyli as foll ia these te cal specificatioas. le ASMN Boil aad Pressure Vessel Code and a licable Add teraLaology r inservice Required f for perfo cies inservice Meekly Monthly~rly or every 3 the Seai ually or every months 9 aunths Year or annually At least once per days At least once per 3 days At least once per 92 ys At least once per'84 ys t least once per 276 s A least once per 366 da 3 h The provisions f Specif icati 1.0.LL applicable to the ahcNe required f encies for r forming ervice testing ivities.4 rmance of the ve iaservice sting ac ivities shall'=e in tion to other s cified survei lance r rements.5.Nothing the ASMB Boile and Pressure essel C shall"e coastrued o supersede the equiremeats o aay te ical specificati 6.The inservice pection progr for piping i tifed i Generic Letter-01 shall be pe ormed in acco dance wx..staff positions o schedule, met, personnel, and sa~expansion included n this generic l tter.BFN Unit 3 1.-12 0 0 0 I~$'fee i0'n,po~l.t<~'2 t995 Sa S utdo o-BFH has deve ed an Appendix R fe Shutdown ogram.This pr am is to ensure at the equipment re red by Appen R Safe Shutdo Analysis is mai ained and demons ated functio as follows: 1.The fun ional requirem s of the Safe utdown Syst and uipment, s well as appro fate compensat measures hould th e syst components be ble to perform heir'nten d junc on are o ined in Section II of the pro am.fiick pgJoad.e.5 PP.~Rahu 2.Testing and monitor of the Append R Safe Shutd systems and equipment are defin in Section V o the Program.c c R,C;c.<o~csi C~i s 0 C-The COLR s the unit-specific for the current document that provides cycle.These cycleKspecific be determined for each oyez'c cat o..Pl addressed in individual s e a4ag cycle in accordance with ant operation within these limits is fications. Nl h LIMITI C COllTROL OD Ph 11 I e a pat rn w resu s in e core b ng on a t ermal imit, i..opera ng on limit val for , LHCR, r M CllSFR1 I (>eyes/TI6mer 2.LI pager/9 WS'ee.r 3 a){Pagcg BFS Unit 3 1.0-124 NENDMNTNtL 2 0 0 PAGE~SOP 0, 0' ~, ACTIONS LEAKAGE INSEFtT<i~Page 1 of 3)Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated conditions within specified Completion Times.LEAKAGE shall be: a~Identified LEAKAGE LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank;or 2.LEAKAGE into the drywell 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;C.d.Unidentified LEAKAGE All Leakage into the drywell that is not identified LEAKAGE;Total LEAKAGE Sum of the identified and unidentified LEAKAGE;Pressure Boundar LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS)component body, pipe wall, or vessel wall.The definitions found in this insert will be placed in alphabetical order with the other ISTS definitions.

INSERT 1 (Page 2 ot 3)P/~LINEAR HEAT GENERATION RATE (LHGR)MODE PHYSICS TESTS SHUTDOWN MARGIN (SDM)The LHGR shall be the heat generation rate per unit length of fuel rod.It is the integral of the heat flux over the heat transfer area associated with the unit length.A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table l.l-l with fuel in the reactor vessel.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 13.10, Refueling Test Program, of the FSAR;b.Authorized under the provisions of 10 CFR 50.59;or c.Otherwise approved by the Nuclear Regulatory Commission. SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: a0 b.c~The reactor is xenon free;The moderator temperature is 68 F;and All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.P~SE~(OF~>

INSERT 1 (Page 3 of 3)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 TURBINE BYPASS SYSTEM RESPONSE TIME THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components: a 0 The time from initial movement of the main turbine stop valve or control valve until 8(C of the turbine bypass capacity is established; and b.The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.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.

INSERT 2 Definitions

1.1 Table

1.1-1 (page 1 of 1)MODES MODE TITLE REACTOR MODE SMITCH POSITION AVERAGE REACTOR COOLANT TEMPERATURE ('F)Power'Operation 2 Startup 3 Hot Shutdown(a) 4 Cold Shutdown(a) 5 Refueling(b) Run Refuel(a)or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel NA NA>212<212 NA t (a)All reactor vessel head closure bolts fully tensioned.(b)One or more reactor vessel head closure bolts less than fully tensioned. BFN-UNIT 2 1.1-7 Amendment

INSERT 3 (Page 1 of 21)Logical Connectors 1.2 1.0 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~A and gg.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'he 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 indentions 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. EXNPLES The following examples illustrate the use of logical connectors. {continued) BFN-UNIT 3 1.2-1 Amendme t PAGa><OF~> 0 INSERT 3 (Page 2 of 21)Logical Connectors

1.2 EXAMPLES

(continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A.LCO not met.A.l Verify...A.2 Restore...In this example the logical connector~AN is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.(continued) BFN-UNIT 3 1.2-2 Amendment ppr..g 4~OF~

1.2 Logical

Connectors INSERT 3 (Page 3 of 21)pi(Logical Connectors

1.2 EXAMPLES

(continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A.LCO not met.A.l Trip.A.2.1 Verify...A.2.2.1 Reduce...A.2.2.2 Perform... A.3 Align...This example represents a,more complicated use of logical connectors. Required Actions A.l, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector Qg and the left justified placement. Any one of thes'e 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 Qg indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed. BFN-UNIT 3 1.2-3 Amendment 0 INSERT 3 (Pago 4 of 21)Completion Times 1.3 1.0 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 aafe 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 Times(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 NODE 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.'hen 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 divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will~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.(continued) BFN-UNIT 3 1.3-1 Amendment PAGE t i 1.3 Completion Times lNSERT 3 (Page 5 of 21)Completion Times

1.3 DESCRIPTION

(continued) H, i.RLbbtdi i i, iyi, p i, 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~i~s inoperability; and b.Hust 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'hat have exceptions that allow completely separate re-entry into the Condition (for each division, 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..." Example 1.3-3 illustrates one use of this type of Completion Time.The 10 day Completion Time specified for Condition A and B in Example 1.3-3 may not be extended.BFN-UNIT 3 1.3-2 (continued) Amendment

INSERT 3 (Page 6 ot 21)Completion Times 1.3 1.3 Completion Times (continued) EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. p ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.Required Action and associated Completion Time not met.B.l Be in HODE 3.B.2 Be in HODE 4.12 hours 36 hours 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 HODE 3 within 12 hours~AN in'ODE 4 within 36 hours.A total of 12 hours is allowed for reaching MODE 3 and a total of 36 hours (not 48 hours)is allowed for reaching MODE 4 from the time that Condition B was entered.If HODE 3 is reached within 6 hours, the time allowed for reaching MODE 4 is the next 30 hours because the total time allowed for reaching HODE 4 is 36 hours.If Condition B is entered while in MODE 3, the time allowed for reaching HODE 4 is the next 36 hours.(continued) BFN-UNIT 3 1.3-3 Amendment ppGp P5 OF~ 0 INSERT 3 (Page 7 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.One pump inoperable. A.l Restore pump to 7 days OPERABLE status.B.Required Action and associated Completion Time not met.B.l Be in MODE 3.~AN 8;2 Be in MODE 4.12 hours 36 hours Mhen a pump is declared inoperable, Condition A is entered.If the pump is not restored to OPERABLE status within 7 days, Condition 8 is also enter ed and the Completion Time clocks for Required Actions B.l and B.2 start.If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and 8 are exited, and therefore, the Required Actions of Condition B may be terminated. 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.0 (continued) BFN-UNIT 3 1.3-4 Amendment ppGL~COF~ INSERT 3 (Page 8 ot 21)Completion Times 1.3 1.3 Completion Times EXAMPLES (((((dd((i i d)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'(continued) BFN-UNIT 3 1.3-5 , Amendment PAGE@~OP~ INSERT 3 (Page 9 of 21)p(g)Completion Times 1.3 EXAMPLES (continued) AM.-3 ACTIONS CONDITION RE(VIREO ACTION COMPLETION TIME A.One Function X subsystem inoperable. A.l Restore Function X subsystem to OPERABLE status.7 days 10 days from discovery of failure to meet the LCO B.One Function Y subsystem inoperable. B.l Restore Function Y subsystem to OPERABLE status.72 hours 10 days from discovery of failure to meet the LCO C.One Function X subsystem inoperable. One Function Y subsystem inoperable. C.1 Restore Function X subsystem to OPERABLE status.+0 C.2 Restore Function Y subsystem to OPERABLE status.12 hours 12 hours (continued) BFN-UNIT 3 1.3-6 Amendment

lNSERT 3 (Page 10 ot 21)Completion Times 1.3 1.3 Completion Times EXAMPLES~3.(ti d)Mien one Function X subsystem and one Function Y subsystem are inoperable, Condition A and Condition B are concurrently .applicable. The Completion Times for Condition A and Condition B are tracked separately for each subsystem, starting from the time each subsystem was declared inoperable and the Condition was entered.A separate Completion Time is established for Condition C and tracked from the time the second subsystem was declared inoperable (i.e., the time the situation described in Condition" was discovered). 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 subsystem was declared inoperable (i.e., initial entry into Condition A).The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met.In this example, without the separate Completion Time, it would be 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.The separate Completion Time modified by the phrase"from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO.This Completion Time allows for an exception to the normal"time zero" for beginning the Completion Time"clock".In this instance, the Completion Time"time zero" is specified as cewencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.(continued) BFN-UNIT 3 1.3-7 Amendment vAGe~~<<M-0 INSERT 3 (Page 11 of 21)Completion Times 1.3-4 CONDITION EXAMPLES EMPlLLL-(continued) ACTIONS REQUIRED ACTION COHPLETION TIHE A.One or more val ves inoperable. A.1 Restore valve(s)to OPERABLE status.4 hours B.Required Action and associated Completion Time not met.B.I Be in HODE 3.~AN B.2 Be in MODE 4.12 hours 36 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.Declar1ng subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.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 Condit1on 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 (plus the extensions) expires while one or more valves are still inoperable, Condit1on B is entered.(continued) BFN-UNIT 3 1.3-8 Amendment go Or

1.3 Completion

Times INSERT 3 (Page 12 of 21)p/f Completion Times 1.3 EXAMPLE ELAN.3-5 (continued) 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.l Be in MODE 3.B.2 Be in MODE 4.12 hours 36 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.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.0 (continued) BFN-UNIT 3 1.3-9 Amendment pAGp~SQF~ INSERT 3 (Page 13 of 21)Completion Times 1.3 EXAMPLES EMPILLM 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.P.3-ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.One channel inoperable. A.1 Perform SR 3.x.x.x.+0 Once per 8 hours A.2 Place channel in 8 hours trip.B.Required Action and associated Completion Time not met.B.l Be in MODE 3.12 hours (continued) BFN-UNIT 3 1.3-10 Amendment

INSERT 3 (Page 14 of 21)Completion Times 1.3 1.3 Completion Times EXAMPLES EXNN.3-6 (ti d)Entry into Condition A offers a choice between Required Action A.l or A.2.Required Action A.l 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.l 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.-(continued) BFN-UNIT 3 1.3-11 Amendment 1.3'ompletion Times lNSERT 3 (Page 15 of 2$)fi(8 Completion Times 1.3 EXAHPLES (continued) ACTIONS~3 7 CONDITION RE(U I RED ACTION COHPLET ION TINE A.One subsystem inoperable. A.l Ver ify affected subsystem isolated.~AN A.2 Restore subsystem to OPERABLE status.1 hour Once per 8 hours thereafter 72 hours B.Required.Action and associated Completion Time not met.fl B.1 Be in NODE 3..2 Be in HODE 4.12 hours 36 hours Required Action A.l 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.l.If after Condition A is entered, Required Action A.l 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.O.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 (continued) BFN-UNIT 3 1.3-12 Amendment

INSERT 3 (Page 16 of 21)Completion Times 1.3 1.3 Completion Times EXAHPLES KMRlEl 3 7..(tl d)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.IHHED IATE Qhen"Iaeediately" is used as a Completion Time, the COHPLETION TIHE Required Action should be pursued without delay and in a controlled manner.BFN-UNIT 3 1.3-13 Amendment Qt'~ lNSERT 3 (Page 0?of 21)Frequency 1.4 1.0 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 Limiting Condition for Operation (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, 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.Example 1.4-4 discusses these special situations. 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 (continued) BFN-UNIT 3 1.4-1 Amendment 0 0 INSERT 3 (Page 18 of 21)Frequency 1,4 1.4 Frequency DESCRIPTION (continued) criteria.SR 3.0.4 restrictions would not apply if both the following conditions are satisfied: a.The Surveillance is not required to be performed; and b.The Surveillance is not required to be met or, even if required to be met, is not known to be failed.EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown)is HODES 1, 2, and 3.X.4-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 interval specified in the 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 NODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not (continued) BFN-UNIT 3 1.4-2 Amendment 0 P 1 INSERT 3{Page 19 of 21)Frequency 1.4~~1.4 Frequency EXAMPLES~4-t t')otherwise modified (refer to Examples 1.4-3 and 1.4-4), 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, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.X AMP 4-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits.Once within 12 hours after>25%RTP 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"~ND").This type of Frequency does not qualify for the extension allowed by SR 3.0.2.(continued) BFN-UNIT 3 C 1.4-3 Amendment t'Pgp~8 OF~

INSERT 3 (Page 20 ef 21)Frequency 1.4 1.4 Frequency EXAMPLES~t-l ti dl"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified'ondition is first met (i.e., the"once'erformance 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.dttqdf.-t 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. q tt dt diff tt qi d~ftt 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 within the"specified Frequency." Therefore, if the Sur'veillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2)interval, but operation was<25K 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 with power>25%RTP.(continued) BFN-UNIT 3 1.4-4 Amendment 0 INSERT 3 (Page 24 of 21)Frequency 1.4 1.4 Frequency EXAHPLES lMPlLl.a-a (t')Once the unit reaches 25K RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency and the provisions of SR 3.0.3 would apply.X P-4 SURVEILLANCE REQUIREHENTS SURVEILLANCE FREQUENCY NOTE Only required to be met in HODE 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 HODE I.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 (plus the extension allowed by SR 3.0.2)interval, but the unit was not in HODE 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 HODES, even with the 24 hour Frequency exceeded, provided the HODE change was not made into HODE 1.Prior to entering HODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.BFN-UNIT 3 1.4-5 Amendment

THIS PAGE IHTEHTIOHALLY LEFT BLAHK BES Unit 3 1.0-12b

NOV Rl lS88 S (Shift)D (Daily)W (We y)5 (Nm Q (Qaszterlg SA (cai-Annnall T (T ly)R (Re Sas)5/9 (5 tart p)1.A P (Prior)At least onc per Lf honrs.t least once er nonaal calendar hoar day (a6 ght to midnight), At east once per days.At 1 t once per.31 Prior to r cor startnp e t applica leo lated pri.-co ch release t less once per 3 son or 92 days.A least e per 6 aonths r 184 days.At 1 t per year or 366 ys.At 1 t once er operatizL cyc~BFI Unit 3 X.a-u 4"E"0M'Q.I g 0

NOV 22 1888.THIS PhGE IHTEHTIOHhLLT BLhHK BFN Unit 3 1.0-13a memserr No.aS 0 I 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES A1 A2 A3 All Reformatting and renumbering is in accordance with the BWR Standard Technical specifications, NUREG-1433. As a result, the Technical Specifications should be more readily readable, and therefore understandable, by plant operators as well as other users.Dur.ng this reformatting and renumbering process, no technical changes (either actual or interpretational) to the Technical Specifications were made unless they were identified and justified. A note was added to Section l.1,"Definitions," in order to clarify that the defined terms will appear capitalized and are applicable thn aghout the Technical Specifications and Bases.This addition is administrative in that it clarifies the use of the definitions throughout the Technical Specifications without changing the intent of any Technical Specification. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.C.1 of Section 1.0 is being reworded and moved to Section 3.0 of the BFN ISTS.This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The movement of this portion of the definition is administrative. Any changes to 1.0.C.1 are justified in the change package for Section 3.0.Not used.A5 The rewording and title capitalization of the"Operable-Operability" definition is administrative because the meaning was not changed.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.A specific change to this definition is changing the"and" to an"or" in"normal and emergency power sources." This is an administrative change because currently the definition along with 1.0.C.2 requires only one source to be operable as long as the redundant systems, subsystems, trains, components, and devices are Operable.Current Specification 1.0.C.2 requirements are incorporated into proposed LCO 3.8.1 ACTIONS for when a diesel or offsite power source is inoperable. Thus, the new requirements are effectively the same as the current requirements. In LCO 3.8.1, new times have been provided to perform the determination of redundant feature Operability. These changes are discussed in the Justification for Changes for LCO 3.8.1.BFN-UNITS 1, 2, 8L 3 pat=~t7 REVISION 0 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A6 A7 A9 The definitions of"Reactor Power Operation,""Startup Conditions,""Hot Shutdown Conditions,""Cold Shutdown Conditions,""Startup/Hot Standby Mode,""Run Mode,"Shutdown Mode," and"Refuel Mode" are incorporated into a"MODES" table (Table l.1-1 of the BFN ISTS)with column headings: Mode, Title, Reactor Mode Switch Position, and Average Reactor Coolant Temperature. This change makes the modes more definitive, which decreases the likelihood of being in more than one mode.This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The title"Rated Power" is changed to"RATED THERMAL POWER." Thi.makes the title more accurately match the definition which discusses the thermal power.The title change and changes to the wording make the definition consistent with the BWR/4 ISTS.The portion of the definition dealing with design power was deleted because it is superfluous to the definition of RATED THERMAL POWER.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational),to the Technical Specifications. The rewording and title capitalization of the"Core Alteration" definition is administrative because the meaning was not changed.The provision added that allows control rod movement with no fuel assemblies in the core cell to not be considered a Core Alteration is less restrictive and is discussed in Comment L3 below.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The definitions of Channel Calibration and Logic System Functional Test (LSFT)were changed.The Channel Calibration definition was clarified to exclude Non-calibratable components. The Non-calibratable components will be included in the channel functional test and source test.The definition of LSFT was revised to remove the requirement to include the sensor and the end device.The end device will be tested during the system functional testing requirements of the affected LCO (e.g., proposed SR 3.5.1.9, which tests to ensure an ECCS pump starts automatically on an initiation signal).Since any of the tests can be credited for performance in parts, as long as the whole channel is tested, it does not matter when the sensor and end device are tested (i.e., with the Channel Calibration, the LSFT, or the system functional test).Thus, the accumulation of both of these changes results in an administrative change.BFN-UNITS 1, 2, 5 3 PAGE REVISION 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A10 All 2 A13 A14 In addition, the word"required" has been added to the Channel Calibration, Channel Functional Test, and the Logic System Functional Test definitions. As a requirement for Operability of a Technical Specification channel, not all channels will have a required sensor or alarm function.Conversely, some channels may have required display function.This is the intent of existing wording, and therefore, the revised wording is proposed to more accurately reflect this intent, consistent with the current licensing basis and BWR/4 ISTS, NUREG-1433. Editorial changes to make consistent with the BWR/4 STS, NUREG-l~i3.During ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the TS.The"Instrument Check" definition is reworded and the title was changed to"CHANNEL CHECK." This is an administrative change because the meaning was not changed.During the BWR/4 ISTS English development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The rewording and title capitalization of the"Dose Equivalent I-131" definition is administrative because the meaning was not changed.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.Nine definitions are added to the BFN Technical Specifications. These definitions were added for consistency with the BWR/4 ISTS.These definitions are used throughout the BFN ISTS and in the current BFN Technical Specifications. The defined terms are used in the LCOs, Surveillance Requirements, and Bases of the Technical Specifications and were defined for the convenience of the users of the Technical Specifications. The inclusion of these definitions are deemed administrative and have no impact on their own.Sections are being added to the Technical Specifications. These additions aid in the understanding and use of the new standard Technical Specifications format and style of presentation. Some'conventions in applying the Technical Specifications to unique situations have previously been the subject of debate and interpretation by the licensee and the NRC Staff.Because the guidanCe in these proposed sections is presented in the BWR/4 Standard Technical Specifications, NUREG-1433, as BFN-UNITS 1, 2,!k 3 REVISION 0 PAGE.OF

JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED approved by the NRC Staff, and the guidance is not a specific deviation from anything in the existing Technical Specifications, these additions are considered to be administrative, The added sections are as follows: SECTION 1.2-Logical Connectors Proposed Section 1.2 provides specific examples of the logical connectors"ANO" and"OR" and the numbering sequence associated with their use.This revision is being proposed consistent with the BWR/4 Standard Technical Specification, NUREG-1433. SECTION 1.3-Completion Times Proposed Section 1.3 provides proper use and interpretation of Completion Times.The proposed section also provides specific examples that aid the user in understanding Completion Times.The proposed Completion Times Section is consistent with the BWR/4 Standard Technical specification, NUREG-1433. SECTION 1.4-Frequency Proposed Section 1.4 provides proper use and interpretation of the Surveillance frequency. The proposed section also provides specific examples that aid the user in understanding Surveillance Frequency. The proposed Frequency Section is consistent with the BWR/4 Standard Technical Specification, NUREG-1433. A15 The rewording and title capitalization of the Minimum Critical Power Ratio" definition is administrative because the meaning was not changed.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no-technical changes{either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.A16 Not used.A17 The rewording and title capitalization of the"Average Planar Linear Heat Generation Rate (APLHGR)" definition is administrative because the meaning was not changed.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.BFN-UNITS 1, 2, 5 3 REVISION 0 PAGE

JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A18 AI9 A20 The"Offsite Dose Calculations Manual (ODCN)" definition is moved to Section 5.0 of the BFN ISTS with some wording changes to make it consistent with the BWR/4 ISTS.This is an administrative change because the definition is being moved to another'section and has no impact on any other definition nor does it change the intent of any Technical Specification. Any technical changes will be justified in the change package for Section 5.0.Not used.The definition of"Venting" was deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is administrative with no impact of its own.A21 A22 A23 The"Site Boundary" definition is reworded and moved to Section 4.0,"Design Features." In Section 4.0, a map depicts the site boundary.This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.This is an administrative change because the definition is being moved, with wording changes, to another section and has no impact on any other definition nor does it change the intent of any Technical Specification. The requirements specified by the definition of"Surveillance" are moved to the BFN ISTS Section 3.0,"Surveillance Requirement (SR)Applicability." The requirements were reworded and incorporated into SR 3.0.1, SR 3.0.2, and SR 3.0.3.This is an administrative change because the requirements are being moved to another Technical Specifications section and has no impact on any other definition nor does it change the intent of any Technical Specification. Any technical change will be justified in the change package for Section 3.0.This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The requirements specified by the definition of"Surveillance Requirements for ASHE Section XI Pump and Valve Program" are moved to Section 5.0,"Administrative Controls." This is an administrative change because it is being relocated, with wording changes, to another section and has no impact on any other definition nor does it change the intent of any Technical Specification. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.Any technical changes will be justified in the change package for Section 5.0.BFN-UNITS 1, 2,&3 REVISION 0 PAGE GP/7 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADNINI STRATI VE CHANGES CONTINUED A24 A25 A26 A27 A28 A29 The requirements specified by the definition of"Appendix R Safe Shutdown Program" (Unit 2 and 3 only)are contained in the existing BFN Appendix R safe shutdown program.This is an administrative change because this definition duplicates the 10 CFR 50, Appendix R requirements. This change has no impact on other definitions nor does it change the intent of any Technical Specification. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The"Safety Limit" definition is deleted because the definition already exists in 10 CFR 50.36 and does not need to be repeated in the Te"hnical Specifications. The use of Safety Limits in the proposed BFN ISTS Section 2.0, Safety Limits, clearly depicts they are the limits below which the reasonable maintenance of the cladding and primary systems are assured, and that violations of the safety limits require plant shutdown and regulatory review.The deletion also maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The"Limiting Safety System Settings (LSSS)" definition is deleted because the definition already exists in 10 CFR 50.36 and does not need repeating in the Technical Specifications. The deletion of this definition also maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The"Limiting Conditions for Operation (LCO)" definition is deleted because the definition already exists in 10 CFR 50.36 and does not need to be repeated in the Technical Specifications. Each proposed LCO clearly depicts the minimum acceptable levels of system performance required to assure safe startup and operation of the facility.The deletion of this definition also maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The"PRIOR TO STARTUP" definition is deleted because it will not be used in the BFN ISTS.The Applicability, Frequency, and Completion Times in the BFN ISTS contain specific plant operation modes and do not need further clarification. The deletion of this definition also maintains the consistency between BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.I The"Operating" definition is deleted because this state of a system does not need to be explicitly defined when considering whether or not the design function can be met.Whether a system is"Operating" or"shut down" does not provide relief concerning Operability requirements. The definition of Operable or Operability is sufficient in this case.BFN-UNITS 1, 2, 5 3 REVISION 0 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADM IN I STRATI YE CHANGES CONTINUED 0 A30 A31 A32 A33 A34 Operability is assumed until the system, etc.is found to be inoperable by failure anytime or during the performance of the Surveillance Requirements at the specified.frequencies. The deletion of this definition also maintains the consistency between the BFN ISTS a;sd the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Immediate" is deleted'because it will not be used in the BFN ISTS.The term"Immediately," however, will be used.The use of this term is defined in Section 1.3 of the BFN ISTS.The dele.'ion of this definition also maintains the consistency between the BFN ISTS and the BWR/4 ISTS.This is an administrative change because the term is being moved from one section to another.The removal of this definition is considered administrative with no impact of its own.The definition of"HOT STANDBY CONDITION" is deleted because it is no longer needed.The Startup Mode or Mode 2 contains the conditions of Hot Standby (<1%power)but does not encompass the intent of the Hot Standby Condition. The intent of Hot Standby Condition is to be reducing power and not increasing power as in the Startup Mode.Hot Standby is a condition that is often hard to maintain for many plants and its use was phased out in later BWR operating plants.Actions will require the plant to be placed in an appropriate Mode or other condition instead of the Hot Standby Condition. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"COLD CONDITION" is deleted.Creation of Table 1.1-1 on reactor modes includes appropriate definitions for modes based on reactor mode switch position, reactor coolant temperatures, and reactor head bolt tensioning. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"SHUTDOWN CONDITION" is deleted.Creation of Table l.1-1 on reactor modes includes appropriate definitions for modes based on reactor mode switch position, reactor coolant temperatures, and reactor head bolt tensioning. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Modes of Operation" was deleted because it will not be used in the LCOs or Surveillance Requirements of the BFN ISTS.The Modes of Operation definition was incorporated into Modes table through the definitions of the positions of the mode switch which determines the BFN-UNITS 1, 2, 5, 3 REVISION 0

JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A35 mode of operation. Terms not used in the LCOs or Surveillance Requirements of the Technical Specifications do not need to be defined.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Primary Containment Integrity" is deleted because of the confusion associated with this definition compared to its use in its respective LCO.All the requirements are spe'cifically addressed in the LCO along with other LCOs in the Containment Systems Section (Section 3.6).The Bases for these LCOs also contain a description of what constitutes primary containment. The deletion of this definition maintains, the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.A36 The definition of"Secondary Containment Integrity" is deleted because of the confusion associated with this definition compared to its use in its respective LCO.All the requirements are specifically addressed in the LCO along with other LCOs in the Containment Systems Section (Section 3.6).The Bases for these LCOs also contain a description of what constitutes secondary containment. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.A37 A38 The definition of"Operating Cycle" is deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The BFN ISTS uses specific times (18 months instead of every Refueling) as designations for Surveillance Frequencies. Terms not used in the LCOs or Surveillance Requirements of the Technical Specifications do not need to be defined.The deletion of this definition'maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Refueling Outage" is deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The BFN ISTS uses specific times (18 months instead of every Refueling) as designations for Surveillance Frequencies. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.BFN-UNITS I, 2,&3 REVISION 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A39 A40 A41 The definition of"Reactor Vessel Pressure" (the term"Reactor Pressure Vessel Pressure" is used in the BWR/4 ISTS)was deleted because it is clearly depicted in the BFN ISTS Bases that it is the pressure measured by the steam dome detectors (Ref.Section 3.4.10).The deletior.of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Transition Boiling" is deleted because it is not used in either the LCOs or Surveillance Requirements. A discussion of MCPR and transition boiling is found in Bases 3.2.2.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definitions of the following terms are deleted because they are not used in either the LCOs or Surveillance Requirements of the BFN ISTS.Channel Instrument Functional Test Source Check (Unit 2 only)Simulated Automatic Actuation Instrument Calibration Trip System Protective Action Protective Function Logic Core Maximum Fraction of Critical Power A42 Some of these terms are encompassed in the definitions of"Channel Check,""Channel Functional Test," and"Channel Calibration." Any changes to Surveillance Requirements are justified in the appropriate Technical Specifications Section change package.The deletion of these definitions maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of these definitions is considered administrative with no impact of its own.The definition of"Functional Test" is deleted because it is not used in either the LCOs or Surveillance Requirements. The definition of"Functional Test" is"the manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water)." These types of tests in the BFN ISTS are called out directly in the Surveillance Requirements (e.g., verify the following ECCS pumps develop the specified flow rate).Post maintenance functional testing is covered by plant procedure and not usually in the Technical Specifications. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.BFN-UNITS 1, 2, 5, 3 REVISION 0 PAGE~OP~W JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES'ONTINUED A43 A44 A45 A46 A47 A48 The definition of"Shutdown" is deleted.A"MODES" table that definitively defines separate modes and eliminates the possibility of being in more than one Mode at a time has been created (proposed Table l.l-l).The Shutdown condition can be either in Mode 3, 4, or 5.This is made clear in the Modes table which shows other criteria that must be met to be in those Modes.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Engineered Safeguards" was deleted because it'.s not used in either the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Reportable Event" was deleted because it is not used in either the LCOs or Surveillance Requirements of the BFN ISTS.The use of Reportable Event is covered in 10 CFR 50.73 and does not need to be defined in the Technical'Specifications. Review of Reportable Events is covered in Section 5.0 of the Technical Specifications. The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Purge or purging" was deleted because it is not used in either the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Unrestricted Area" was deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The definition of"Gaseous Waste Treatment System" was deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.BFN-UNITS 1,.2,&3 10 REVISION 0 PAGE

JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION ADMINISTRATIVE CHANGES CONTINUED A49 A50 A51 A52 The definition of"Members of the Public" was deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of, this definition is con idered administrative with no impact of its own.The definition of"Limiting Control Rod Pattern" was deleted because it is not used in the LCOs or Surveillance Requirements of the BFN ISTS.The deletion of this definition maintains the consistency between the BFN ISTS and the BWR/4 ISTS.The removal of this definition is considered administrative with no impact of its own.The Surveillance Frequency Notation Table is being deleted because the Surveillance Requirement Frequencies in the BFN ISTS do not use notation.The Frequencies are specific by giving the number of hours,.days, or months (e.g., instead of M the BFN ISTS will have 31 days).The change of the"Daily" SI frequency from once per normal calendar day (midnight to midnight)to once per 24 hours has been categorized as administrative since over the long-term the frequencies will be about the same.However, the change could be considered more restrictive and less restrictive. It is more restrictive since CTS would allow up to approximately 48 hours (12:Ol am on one day to 11:59 pm on the next day).However, the next day you could only go to ll:59 PM and would likely go less and the previous day you would have likely gone less than 24 hours.It is less restrictive since the new Frequency would always allow as long as 30 hours (24+25%).Over the long-term, on the average the CTS and ISTS frequency will be about the same.The rewording and title capitalization of the"Core Maximum Fraction of Limiting Power Density" definition is administrative because the meaning was not changed.During the BWR/4 ISTS development, certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. This change maintains the consistency between the BFN ISTS and the BWR/4 ISTS.TECHNICAL CHANGES-MORE RESTRICTIVE Not Used.M2 The proposed Startup Mode will now include the mode switch position of"Refuel" when the head bolts are fully tensioned (footnote (a)).The change eliminates the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist.Currently, CTS 1.0.M allows the plant to be considered in the SHUTDOWN CONDITION and BFN-UNITS 1, 2, 8E 3 REVISION 0

JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION TECHNICAL CHANGES-MORE RESTRICTIVE CONTINUED in the Shutdown Mode with the mode switch in the Refuel position (and other positions are allowed while in the Shutdown Mode)as permitted by notes to that definition. By defining this plant condition as STARTUP MODE, sufficiently conservative restrictions will be applied by applicable LCOs.The allowance to place the Mode Switch in other positions has been moved to Section 3.10, Special Operations and Section 3.3.2.1, Control Rod, Block Instrumentation. Any technical changes to these allowances will be discussed in the Justification for Changes to these Sections.TECHNICAL CHANGES-LESS RESTRICTIVE Ll The words"or actual" in reference to the injected signal have been added to the definition of CHANNEL FUNCTIONAL TEST.Some CHANNEL FUNCTIONAL TESTS are being performed by insertion of the actual signal into the logic (e.g., rod block interlocks). For others, there is no reason why an actual signal would preclude satisfactory performance of the test.Use of an actual signal instead of the existing requirement, which limits use to a simulated signal, will not affect the performance of the channel.OPERABILITY can be adequately demonstrated in either case since the channel itself cannot discriminate between"actual" or"simulated." L2 Since the requirements are essentially the same, the analog and bistable channel requirements have been combined.The only technical difference is the location of the injected signal.As provided in the definition of CHANNEL FUNCTIONAL TEST for instruments with analog channels, the signal used to test bistable channels is proposed to be allowed to be injected"as close to the sensor as practicable." Injecting a signal at the sensor would in some cases involve significantly increased probabilities of initiating undesired circuits during the test since several logic channels are often associated with a particular sensor.Performing the test by injection of a signal at the sensor requires jumpering of the other logic channels to prevent their initiation during the test, or increases the scope of the test to include multiple tests of the other logic channels.Either method significantly increases the difficulty of performing the surveillance. Allowing initiation of the signal close to the sensor provides a complete test of the logic channel while significantly reducing the probability of undesired initiation. In addition, the sensor is still being checked during a CHANNEL CALIBRATION. BFN-UNITS 1, 2, EE 3 12 REVISION 0 PP,GOOF 0 JUSTIFICATION FOR CHANGES SECTION 1.0-USE AND APPLICATION TECHNICAL CHANGES-LESS RESTRICTIVE CONTINUED L3 This change is proposed to allow control rod movement in a defueled cell to not be considered a CORE ALTERATION. In this configuration, the negative reactivity inserted by removing the adjacent four fuel assemblies is significantly more than any minimal positive react.vity inserted during the removal of the control rod.Appropriate Technical Specification controls are applied during the fuel movements preceding the control rod removal to protect from or mitigate a reactivity excursion event.After the fuel has been removed, sufficient margin and design features (the design of a control rod precludes its removal without all fuel assemblies in the cell removed)are in place to;";low removing the Technical Specification controls during the control rod removal.The proposed change focuses the definition on activities that can affect core reactivity. Maintaining CORE ALTERATIONS as movement of only that which can affect core reactivity is consistent with the BWR Standard Technical Specifications, NUREG 1433.The basis for this is evident in that the Specifications that are applicable during CORE ALTERATIONS are those that protect from or mitigate a reactivity excursion event.BFN-UNITS I, 2, 5 3 13 REVISION 0 PAGE/3 pF yg

UNIT<CURRENT TECHNICAL SPECIFICATION MARKUP 0

2.1 Applies

o the interrelated variabl s associated with f 1 therma behavior.Appli to t p~ettings of the'rum ts and devices wh'are ovided to revent t e react system s ety imits f om heing e ceded.To establish limits w ich ensure the integrity of the fuel cladding.To efine the evel of the pr cess vari les at which a tomatic p tective action s initiat to prevent the fuel clad g integrity safety.li t from being ceded~,o see q a~;>s (sc s)2e(Sl s Ical hr Porc~45 The limiting safety system settings shall be as specified below: 'Cgcgff l7l4rgW tvi'H pkC P eaetor)reiaure~<~,.~mLa-aacf gore Plo~IL reac'tor'ressure is eater 8 0 psi, the sten 0 of minimum c tical wer atio.()le than 1.10 shal cons itute viola ion of the fuel cladding integrity safety limit.N cf'R 5~II L c K l s I 6 a~When the Mode Switch is in the RUN position, the APRM flux scram trip sett;;.g shall be: APRM Flux Scram Trip Setting (Run Node)(Flow biased)BFN Unit 1 1.1/2.1-1 0 0 gLs)2.l.h d.Fixed High Heutron Flux Scram Trip Setting-When the mode svitch is in the RUE position, the hPRM fixed high flux scram trip setting shall be: Sgl20X pover.L4dh the 5+aen do 70$cacto ress re o rPorogloo g, O~aeoO.Core loA en the reactor pre sure i F800 p ia or core lov is 10X of rated the ore the 1 pov r sha 1 no exce 823 t (.2 X of rated erma pover.TRH2mal Power SholL be~PM/u P.Tf'.2.APRM and IRM Trip Settings (Startup and Hot Standby Modes)~a.hPRM-When the reactor mode svitch is in the STARTUP position, the hPRM scram shall be set at less than or equal to 15K of rated pover.b.IRN-The IRM scram shall be set at less than or equal to 120/125 of full scale.e~~+<'<5>~4r Chz~g5 A gpss lsT5 r3,l,l BPK Unit 1 lol/2.1W PAGE~CPM 0 5 cc'p LI I To urc t t thc AFETY IMITS est lishe in Sp ifica ion 1..h are ot ex eded, each r uired scram all b itiat d by i s expc ted cram signal The hFEIT IMIT hall be as cd t be ex cede when scr is ac omplis ed b means other than the ected scram signal..1.B.2~3~Scram-turbine stop valve closure g 10 per-cent valve closure Scram-turbine control valve fast closure or turbine trip g 550 psig Scram and isola-g 538 in.tion (PCIS groups above 2,3,6)reactor vessel lov vater level zero 4.(Deleted)5.Sera~main S 19 percent stcam linc valve isolation closure.6~Main stcam g 825 psig isolation valve closure-nuclear system lov pressure C.cv ther is ir diated fue in e re tor v sel, c ter evcl 1 b gr atc or e to 372.inches bove easel ero.'.W Core spray and LPCI actuation-reactor lov vater lcvcl 398 in.[above vessel zero~/e l<3 gcRc4)f Qgg fg L Lvg~, Mtl~(re~then H,<Wt opme<CH~C i~vo4~yg Qq], 2~3~HPCI and RCIC actuation-reactor lov vater level Main steaa isolation valve closure-reactor lov vater level 470 in.above vessel zero s9s in.I above vessel zero gc'e 3w+Acahon 4i 5g~9 8Fhl tsT5$.3,f.f y,g,g.i, 3'.3s 5'.2-And 3,3.4, (BFN Unit 1 1.X/2.1-5 1.2/2.2 R YS INT ITY SAFETY LINZT(1.2 Reactor Coolant S s a Inte r LI TIN3 SA Y SYSTHPI I K 2 2 Rea tor Coolant steII Inte it A 11 bllit A licabillt Appl es to limit on rcac r coolant sys em pressure pplies to t ip settings of the instruments and devices which arc provi d to preven the reactor stcji safety lmlts fraa be g exceeded.0~beet ive To establ sh a limi below wh ch the inte rlty of t e reactor coolant systea is t threa cned due to an overpressure condition. Ob ec ive To efine the lev 1 of the pr ess variable at which a cxaatic prote ive action is initiated to prevent thc pressure safety limit fraa being exceeded.S ccificatlons 5'L a.i.>e press re at the lowe t point o the rea tor ves el sh 1 not exc ed 1,37 psig encvc irra ated f 1 is i the reacto vesse gmc4<g~dome pc~ce'~(sos s" A.Nuclear systea relief valves open-nuclear system pressure 1,105 pslg+ll psl (4 valves)1,115 psig+ll psi (4 valves)The lhaiting safety systea settings shall bc as specified below: Liraiting Safety Protective Action S stem Settin ee~st',Razhon 4r 5Ignys+S~<t~T's 3.z,l.l a,ds.q,3 1,125 psig+11 psi (5 valves)B.Scraa--nuclear <1,055 psig systcra high pressure HPH unit 1 1.2/2.2-1 7S z7g Cf.'P CGgg'6363~6.3 Each member o t e uni f h it's staff shall meet or exceed the minimum qualif Scations or camp i f mparable positions as specifi.ed in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN69-A). 6.4{Deleted)6.5 (Deleted)6.6 (Deleted)6.7.1 e following ctions shall be t en in the e v t a Saf ety Limit is violated: f/a The Operations Cent shall be not fied by tclepho as soo as possible and i all cases wi 1 hour.The Site Vi e President and t NSRB shall notified vithi 24 hours.b.A Safety Limit Vi lation Repart shall be prepar.Thc report shall be revie by the PORC This report rt s ll describe{1)applicable cir tances prc eding the viol ion (2)effects of the viola ion upan faci ty components, systems, or structures, and (3)corre tive ac tion t to prevent recurrenc s ll be submitted to the c.The Sa ety Limit Vio+tion Report s Coned.ssion, the NSRB, and the Site ice President within 14 days of the violation. /f<oPS6d 2.2.PAGp~pF~BFN Unit 1 6.0-5 0 itical operation of t unit shall not e resumed until author by the Commission. 6.8 6.8.1 PROCEDURES Scc duple (icr,f~~for C&gcj 4y~)gal I g7 J 5 (3 6.8.1.1 Written procedures shall be established, implemented ard maintained covering the activities referenced below: a.The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.b.Limitations on the amount of overtime worked by individuals performing safety-related functions in accordance with NRC Policy statement on working hours (Generic Letter No.82-12)c.Surveillance and test activities of safety-related equipment.(Deleted)e.(Deleted)f.Fire Protection Program implementation. g.(Deleted)h.(Deleted)i.Offsite Dose Calculation Manual.)~Administrative procedures which control technical and crcss-disciplinary review.BFN Unit 1 6.0-6 PAGE 7 gj~ 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP pAGE~OF~ 0

2.1 Applie

to the inter elated vari les associate with fuel the 1 behavior.Applies to rip settings of the inst ents and dev ces which ar provided to revent the rea or system s fety limits from being e ceeded.T establish l'mits which sure the integrity of the fuel cladding.To define the evel of t e p ocess vari es at w ch tomatic pr tective ction s initiate to pre t the fuel cladd ng int ity safety limit from being exceeded.v,o sneezy c-z~'(sos)>(S~s gee(44 Grc s~Q.J.(I C$<rj r hS l7I~yegg p essure is gr ater 800 sia, he exi tence of a nimum crit al er ra io (MCPR less han 1.10 ll c nstit e vi latio of th fuel cia ing integrity safety limit./H<<R,~4((kc.~I,io o c~~C Reac:tor Psssme ps'nd Pore Plo~10%of pate dl When the reactor The limiting safety system settings shall be as specif ied below: l.APRM Plux Scram Trip Setting (RUN Node)(Plow Biased)a.When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be: SC<Su&i f~'In Qq<~CS+0 8W~15 f 5 Z.3 I BFH Unit 2 1.1/2.1-1 WGE 0 4~,~~~~a$1 2.0 SAFETY LIMITS CS4s 2~SpectliQio~ g.0 MAR OST 88 1.A u o~(Cont'd)B.l I.l J~c.gft0t E5'/reactor ressure~~or gare g'lov OX ettt rare~Wh the r actor res ure s g 0 ps or c re ov i gl of r ted, e c r th al over hall ot exc d 82 MWt 5X, rated thermal po er).7MeshC.Pe~ca<L~ll ge.-2<Yo RVP..d.Fixed High Neutron Flux Scram Trip Setting-When the mode svitch is in the RUN'osition, the APRM fixed high flux scram trip setting shall be: SS120'I pover.2.APRM and IRM Trip Settings (Startup and Hot Standby Modes).a.APRM-When the reactor mode svitch is in the STARTUP position, the APRM scram shall be set at less than or equal to 15K of rated pover.b.IRM-The IRM scram shall be set at less than or equal to 120/125 of full scale.~+<kali CCki4~gl~>><I<<tsar z.g.i.l BFN Unit 2 1el/2.1-4 AMENDMEMT NO.Z43 PAGE

SAFETY LIMITS C SLc IMITIHG SAFETY SYSTEM SETTIHG 1.1.B.v.1.B.ovc Sett s ensure t t the Sa ety Limits est lished i Specific ion 1.1.are not ex cded, eac required scr shall be itiated b its expected ram signa e Safety L mit shall e assumed to e exceede vhcn scr is accomp ished by means other th the expected scram sig al.2~3~Scram and isola-2 538 in.tion (PCIS groups above 2,3,6)reactor vessel lov vatcr level zero S cram-turbine stop valve closure g 10 per-cent valve'losure Scram-turbine g 550 psig control valve fast closure or turbine trip 4.(Deleted)5.Scram-main g 10 percent steam line valve isolation closure 6.Main steam 2 825 psig isolation valve closure-nuclear system lov pressure C.W v Se QA~t enever ther is ir adi ted el i the re ctor v se t vate level hall b grc ter t or e al to 372.inche above essel zero Z.l l.g lPcac,kcr yc5gg~kc~<<<s4(J Q)reefer~<~~<p aFR z,<cd.v4.ilrad,qgcf gag,L 2.HPCI and RCIC actuation-reactor lov water level 470 in.above vessel zero 3.Main stcam isolation valve closure-reactor lov vater level 398 in.above vessel zero 1.Core spray and Z 398 in.~LPCI actuation-above reactor lov vessel vater level zero Se~<SL;4'-.Ck~ Q, (g~~~N'<~<<~~(, S.Z.S.I, 8.3.W.2-~3.~.C.i BFN Unit 2 1.1/2.1-5 NENOMEHT NO.I 8 3 pAGE 0 0~ 1.2/2.2 REACTOR LANT SYSTEM INTEGRITY SAFETY LINIT5 Lg 1.2 Reactor Coolant S stem Inte r LINITING SAFETY SYSTBH SETTING 2.2 Reac r Coolant S stem Inte rit A cabilit A licabilit pplies to limits on r ctor coolant system pressure.Applies to trip s tings of the instruments and evices Mich are provided t prevent the reactor syste safety limits from being ceeded.o~b ective To establ h a limit hei~Mi the inte rity of the reactor coolan system is not threa ned due to an overpressure cond tion.Ob ectiv To de ne the level"the pr ss variables a which au omatic protect e action is initiated to revent the pressure safet limit from being exceede ecifications SL Q.I.N e press re at the~st point o the rea or vessel shal not exc d 1.37 sig Men ver irrad ted fu is in th reactor vessel.e Ju ephor(rsas p.~.The limiting safety system settings shall be as specified belch: Limiting safety Protective Action S stem Settin A.Nuclear system 1,105 psig+relief valves ll psi open--nuclear (4 valves)system pressure'.1,115 psig+ll psi (4 valves)~IRXg4/i40~ 4v~~4 EFN IS&ggf./a~gg3 1.125 psig+11 psi (5 valves)B.Scram-nuclear<1,055 psig system high pressure BFN Unit 2 1.2/2.2-1.".AGE 5=2. 0 6.3 8 72.Scc h'og Each member of the unit's staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A) .6.4 (Deleted)Jw~hf c'R4+A)f O~g+J/'PnJ J 5 T5 6.5 (Deleted)6.6 (Deleted)6.7 6.7.1 The following ctions sha 1 be taken i the event a Safety Limit/'s violated: a.The Operatio s Center shall notified y telephone as/soon as possibl'nd in all cas s within 1 our.The Site/Vic'resident and the NSRB s 11 be noti ed within 24/h s.I b.Safety Li't Violation Re rt shall prepared.The I report sha be reviewed b the PORC.This report s 1/describe)applicable c rcumstance preceding the//!violatio , (2)effects o the viola on upon facil y i compon ts, systems, or structures, and (3)corre tive/acti taken to preve recurrenc/c.Th Safety Limit V lation Repor shall be s itted to the C mmission, the N RB, and the S te Vice Pre ident within 14 ys of the vio ation.I d.Critical ope ation of the t'hall no be resumed until authorised y the Commissi BFH Unit 2 $.8 6.8.1 PROCEDURES Qe<34d+PicW~~C~gs+~@Pal)$75 5;n 6.8.1.1 Written pzocedures 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.Limitations on the amount of overtime worked by iraividuals performing safety-related functions in-accordance with NRC Policy statement on working hours (Generic Letter No.82-12).c.Surveillance and test activities of safety-related equipment. d.(Deleted)e.(Deleted)f.Fire Protection Program implementation. g.(Deleted)h.(Deleted)i.Of fsite Dose Calculation Manual.j, Administrative procedures which contzol technical and cross-disciplinary review.BFN Unit 2 6.0-6 PAGE~OF~ 0 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP

1.1 Applies

o the interr lated variabl s associated with fuel therma behavior.r A lies to tri settings of e instrumen and devices hich are pr ided to prevent the reactor ystem safety limits fro being exceeded.To establish li its which e ure the int rity of the fuel cladding.(Q g.o sstF~q w pg (S 2.1 5 Lg 4cac~Car<Scs To def e the level of the proce s variables t which aut tic protec ve action is itiated to prevent the fue cladding'egrity safety limit om being eded The limiting safety system settings shall be as specif ied below>g y,'fA fAc K~~c ReaereJjressere ye4a and gore Plo Qsree er par~~<>Ra!~a APRM Plux Scram Trip Setting (Run Mode)(Plow Biased)eygg4ao~wyeg when the reactor pressure is greater than 800 psia, the existence of a minimum.critical power ratio (NCZ%)less than 1.10 shall constitute violation of the fuel cladding integrity safety limit.mcfl sgll W~l.to a elIhen the Mode Switch is in the RUN position,"he APRM flux scram trap setting shall be: 5'~MQ:~on+g~~II8$40 Pph)Pf$3'Ze J,/BFN Unit 3 1.1/2.1-1

~@'.5c 2.1.k MAR 09 1999 d Fixed High Hcutron Flux Scram Trip Setting-Wh thc mode svitch is in t e RUB position, the APE fixed high flux scram trip setting shall be: Sgl20X povcr.N;fh H'+~+m FS reactor Pressure yaka<Ps'g or Jforeglow 0 of rate.arr Cgak APRN and IRM Trip Settings (Startup and Hot Standby Modes).epopt a~war en the reactor prc sure F800 p a or re lov is 10K of ated, the orc the pov shal not cxce 823 (.25 of rated crmal over)AC<<4L Posses Shall be'05%RTP ao kPRN-When the reactor mode switch is in the STARTUP position the hHN scram shall be set at less than or equal to 15%of rated pover.*b.IR5-The IRN scram shall be set at lcaa than or equal to 120/125 of full BFK Unit 3 AMBlgggf7 gg Z g 8 0' SLs sPeci Red)gn Q.o~JUL 17 1 1.B.To urc that S LI TS cata ished in ecifi tion 1.1 h are not ccedc ea rc uircd scr shall e i itiated by ts exp cted cram s gnal.Th SAFETY LIMIT hall e assumed to bc ceeded vhcn scram is ccompl hcd by cans other thc ected cram signal..1.B.2~3~Scram-turbine stop vLlve closure Scram-turbine control valve fast clouds c-turbine trip g 10 per-cent valve closure g 550 psig l.Scram and isola-g 538 in.tion (PCIS groups above 2,3,6)reactor vessel'ov vatcr level zero 4.(Deleted)5 Serac~in S 10 percent steam line valve isolation closure 6.Main ateaa 2 825 paig isolation valve closure-nuclear ayatca lov pressure eneve th re ia irra atcd z 0~~I t3 8mcbr VCmI.mm~S)I bc gr<siM thin Qc Hf PF the aCH~irrad ice~W el in e eacto vea 1, th vate lev bc gre er o equal to l 372.inche abo e vea 1 20 30 Core spray and LPCI actuation-reactor lov vater level HPCI and RCIC LctuLtion reactor lov vater level Main steam isolation valve cloaure-reactor lov vatcr level 398 in.Lbovc vessel zero 470 in.above vessel zero 398 in.Lbovc vessel zero 5e<A~t'Wysn gr&~c5 y DF=ll IS75 7,3,I.I,$.3,S,I 3.3 g.g~g X 3(.l BPÃUnit 3 1 ll2.1-5 0 (S<o)1.2 2.2 Applies to imits on reac r coolant syatea p essure.kppl es to trip set ings of the i raments and d ices vhich e provided to revent the eactor syatea afety limits froa being ex ceded.To establis a lia1t belov which the integ ty of the reactor coolant stea is not threatened due to overpresaure conditi To def e the level of the proce variables at vhich aut tic protective action i initiated to prevent the esaure safety limit froa being exceeded.SL g,i~e pres e at e lou t point of the rea tor ve el 1 not exc 1,3 paig eneve irra ated f 1 is the reacto veaae@<<4<S~J~e pi<ssune~tl be.a ip~g The limiting aaf ety'yst col settings shall be as specified helot: k.Nuclear systea 1,105 psig g relief valves ll pai open-nuclear (4 valves)system pressure 34Es+Pcag~ <gg~8W><AS~z~~<...1,115 craig g 11 pai (4 valves)1,125 peig g 11 psi (5 valves)B.Scram-nuclear Z1,055 psig systea high pressure BPS Unit 3 3..2/2.2-1 7$+72 ac&'c n 2 i o 6.3 Each member of the unit's staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A) .6.4 (Deleted)r<~'hg>'ca Q'om@p g/+8 FN]g yg 5.D 6.5 (Deleted)6.6 (Deleted)6.7 6..1 The f llowing aetio s shall be taken in the event a Safety Limit is v olated: a.The NRC Ope tions Center s ll be notified by telephone as soon as po sible and in al cases within 1 hour.The Site Vice Pres dent and the N shall be not ied within 24 hours.b.A Saf y Limit Violat on Report shal be prepare.The/repo t shall be rev wed by the POR.'his re rt shall de ribe (1)appli able circumstan es prece the v'olation, (2)e ects of the vio ation upon facility/omponents, sys ems, or structur s, and (3)correctiv action taken prevent recurrence. /The Safety imit Violation R port shal be submi ed to the Commission the NSRB, and t e Site Vi e Presid t within 14 days of t e violation. I d.Criticy operation of t e unit s ll not be resumed, until authogized by the Commission. BFN Unit 3 6.0-5 PAGE OF 0 0

6.8.1 PROCEDURES

6.8.1.1 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.Limitations on the amount of overtime worked by individuals performing safety-related functions in accordance with NRC Policy statement on working hours (Generic Letter No.82-12).C~Surveillance and test.activities of safety-related equipment. d.(Deleted)e.(Deleted)f.Pire Protection Program implementation. g.(Deleted)h.(Deleted)Offsite Dose Calculation Manual.j.Administrative procedures which control technical and cross-disciplinary review.BFN Unit 3 6.0-6 PAGE~iQF~ 0 t 0 JUSTIFICATION FOR CHANGES SECTION 2.0-SAFETY LIHITS AOMINISTRATIVE CHANGES A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which re ulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 The reactor pressure limit unit of measure has been changed from psia to psig.In addition, the requirement for when the HCPR limit is applicable has been reduced slightly (by adding the"equal to" sign)for consistency with the BWR Standard Technical Specifications, NUREG-1433. The limit on core flow is now specified as greater than or equal to.The current Safety Limits do not address the situation when core flow is equal to the limit.While these changes are actually more restrictive, since they are so minor, they are considered an administrative changes.A3 The Safety Limits were reworded without changing the intent of the Safety Limit (no technical changes were made).Editorial rewording is consistent with the BWR Standard Technical Specification, NUREG-1433. During its development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the TS.Units for pressure has been changed from psia to psig.The Safety Limit was changed from pressure (1375)at the lowest point of the reactor vessel to reactor steam dome pressure (1325), which is equivalent considering water level differences. BFN-UNITS 1, 2,&3 I p><e;~pi=~Revision 0

JUSTIFICATION FOR CHANGES SECTION 2.0-SAFETY LIMITS A4 The"equal to" was taken out of"less than or equal to" symbol.This was done for consistency with the current BFN Bases for the Safety Limit which states that a core thermal power limit of 25 percent for reactor pressures below 800 psia (785 psig)is conservative. This is also consistent with NUREG-1433. Also the"equal to" was taken out of"less than or equal to" symbol as it relates to rated core flow to ma-:ntain consistency between the current technical specifications and NUREG-1433. TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new 2.2 requirement is added to the Safety Limit Violations Sec.'ion, which requires all SLs to be restored and all insertable rods inserted within 2 hours.Exceeding a Safety Limit may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100 limits.These requirements ensure that the operators take prompt remedial action and also ensure that the probability of an accident occurring when a Safety Limit is violated is minimal.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl BFN proposes the requirements related to Safety Limit Violation reporting and restart authorization not be retained in Technical Specifications. Duplication of the regulations provided in 10 CFR 50.36, 50.72 and 50.73 is not necessary to assure safe operation of the facility.The current regulations require BFN to perform all the actions currently required by Technical Specifications. This change is consistent Technical Specification Change Traveler TSTF-5 (approved by NRC on Il/27/95)and Revision 0 to Generic Change BWROG-09, which addressed several NRC and Industry initiatives to improve the content and presentation of Administrative Controls.BFN-UNITS 1, 2, 5.3 Revision 0 PAGE 0'

1.0 Sftci4icajon

3.p SCC'45%kiCRhOA g c~~+S~.f ST5 Scc~n f,o Ths aucce requently ua are czp c t y defined so that a unifora interpretation of the ayecLficationa may be achieved.reasonable aaintcnance of the cladding and primary systems are augured.Exceeding such a limit requLres unit shutdown and reviev by the Atoaic Energy Coaalsaion before resIarption of unit oyeration. Oyeration beyond such a limit may not Ln itself result Ln serious consequences but it Indicates an operational deficiency aub]ect to regulatory review.-The limiting safety system settings are settings on Lnatrmcntation which initiate the autoaatic protective action at a level such that the safety limits vill not.be exceeded.The region between the safety limit and these settings represents margin vith normal oyeration lying belov these settings.The margin has been established so that vith proper operation of the Instrmcntation the safety liaLta vill never be exceeded.C.-The liaLting conditions for operation syecify the ainiaaa acceytable levels of ayatea performance ILeceasary to assure safe atartup and operatioIL of the.facility.Mhaa these conditions are met, the ylant can be oyeratcd safely azLd abnormal.situations can be safe+controlled. LI kgb Vlf'~pose.'4 LCO 3.O,3, Zn e event a Liaiting Condition for Oyeration'r associated requircaenta. Cannot be satisfied ecauae of cLr tancea In cess of se addr sed the ayecifi tion,'he t shel be ylac in at east t Standby thLIL 6 ho ma in old Shut vi f loving hours unl s corr tive ae ea ar c leted t p rait o ration er the ermiaai le disc very or ti the r ctor is placed an oper ional condi on in ch sy ificati is not ayplicab Except to e equir enta a 1 be st ted in.individ speci ca ons.a yro dea act ons to b taken fo ,circeaa c not rectly ovided r Ln th yyecifica one and e occurrence vo viola the tent ot e specif cat on.For exampl if a'p ificati calls f r syateas (or au ayat to be o able and rovides r explicit requirca ts I5, ne ays (or au stem)ia erable, then Ig th terna or bsyatcms ,are inope ab e the unit ia to be in t 1 ast ot Stand in 6-hours ln Cold Shutlown vi fo oving 30 urs if the Inoperable condition is not cor ected.BF5 Unit 1 On L(o uo (mm Roo ico Roo c.co boo~R3 Roo~R~Roo~co 3,of'eo 2 7,0+'.o,5 3,o,k(g,p,7 1.0-1

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 JUSTIFICATION FOR CHANGES SECTION 2.0-SAFETY LIMITS"Specific" Ll The proposed change deletes the"Power Transient" Safety Limit.The intent of this Safety Limit was to ensure that other Safety Limits are not exceeded.This Safety Limit is assumed to be exceeded when a scram is accomplished by means other than the expected scram signal.The scram setpoints are established in order to ensure margin to the safety limits.Exceeding the scram setpoint, in and of itself, does not necessarily indicate that a Safety Limit has been exceeded.Section 2.1.B,of the present BFN TS contains six power transient trip settings that initiate a reactor scram.These scram setpoints are included in Table 3.3.1.1-1 of the new ISTS.The surveillance requirements imposed on these scram setpoints in Table 3.3.1.1-1 help to ensure that the margin to a safety limit is preserved. The redundancy built into the RPS system is maintained by the action provisions of ISTS 3.3.1.1.Therefore, the intent of present Power Transient Safety Limit 1.1.B is maintained by the proposed provisions in ISTS 3.3.1.1 for the RPS.Additionally, although the proposed changed deletes the requirement for assuming the Safety Limit is exceeded when scram is accomplished by means other than the expected scram signal, the proposed change does not preclude the required actions if the Safety Limit is actually violated.The current Safety Limit for the reactor vessel water level is that level shall be maintained not less than 372.5 inches above vessel zero.This proposed Safety Limit is that level should be greater than the top of the active irradiated fuel (approximately 366 inches above vessel zero).This represents a less restrictive change since the top of the active irradiated fuel at BFN Unit 2 is less than 372.5 inches above vessel zero.The change still ensures adequate margin for effective action in the event of a level drop.This change is consistent with NUREG-1433. BFN-UNITS 1, 2, 8.3'-------'evision 0::

1.0<<~~~~<~ho~Qv (~~<+(5T~Srcoon),o SPt.C,'@cahot r P,O (Con'd)owned, leased, or otherwise controlled by TVA.-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 industrial, commercial, institutional, or recreational purposes.ZZ.-The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in pCi/gm)which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, Z-132, I-133, Z-134, and I-135 actually present.The thyroid dose conversion factor used for this calculation shall be those listed in Table ZIZ of TZD-14844"Calculation of Distance Factors for Power and Test Reactor Sites".-The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.-An individual in a controlled or UNRESTRICTED AREA.However, an individual is not a MEMBER OF THE PUBLIC during any period in which the individual receives an occupational dose (as defined in 10 CFR 20).Qp4cc>,'~Prepo~sR g.o.(Qfke uKtk~Pro po5+gp 3.0rz.S illance quirements shall be et d xqg t e OPE TZONAL NDZTIO or othe condition specif ie f or ingividual limit g condi ons fo operatio unles erwise s ted divi al Surveillance ireme a ex ance Requir men 8 er orme ed ei ance i erval ith a maximum lowable xten ion no to exce 25 ercent the eci ied surveil ce int al.Zt is ot int ded t this (exten ion)rovis'on be us d rep atedly as a co enie e to e end s eil ance inte als bey d t spec'ed for surveillan es at are performed d ing r fueling outage e fo e of a Surveill ce Requiremen within the speci ied time t al s l con itute c liance and PERABILITY equir ents for'i ting ndition for oper tion and ass ciated acti n sta ements uales othe se requ ed by t se specific tions.S eilla ce'r ents d not hav to be p rformed i o erable e'e t.~30'f it is discovered that a Surveillance was not performed within its specified frequency, the 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 less.This delay period is permitted to allow performance of the Surveillance. /)DO 5 5 BFN Unit 1 1.0-11 0 0 ($27$5/CIVIC.~/l~ 3 0 N 3:0.3, ((.'onk.)If the surveillance is not perf ormed within the delay period, the LCO must iagagdiately 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.MM.Surveillance Requ en Section XI Pump and Valve Program-Surveillance Requirements for Inservice Testing of ASMB Code Class 1, 2, and 3 components shall be applicable as follows: Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI.of the ASME Boiler and Pressure Code and applicable Addenda as required by 10 CPR 50, Section 50.55a(g), except where specific written relief has been granted by the Ceaaission pursuant to 10 CFR 50, Section 50.55(g)(6)(i).Surveillance intervals specified in Section XZ of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications: ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice Required f requencies for performing inservice Sei ZwggIc,l~*~CWgg~g>sr~Is~sS weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once'per 184 days At least once per 276 days At least once per 366 day 3.The provisions of Specification 1.0.LL are applicable to the ahem>>required frequencies for performing inservice testing activities. Performance of the above inservice testing activities shall"e in addition to other specified surveillance requirements. 5.Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical speciiication. BFH Unit 1 6.The inservice inspection program for piping identified in'..r.Generic Letter 88-01 shall be performed in accordance wx:."..-.staff positions on schedule, methods, personnel, and samp.'xpansion included in this generic letter.PAGE~O"~1.0-12 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP ~, 1.0 c W<+igl gg4o~for+a ggN I$TS Sccfiou (.ti~The succeeding frequently used terms are erplicftly defined so that a uafform fatcrprctatfon of the specfffcatfoas mny be achieved.reasonable mafatensace of the claddin'g and primary systems are assured.Ezceeding such a limit requires unit shutdovn and reviev by the Ltomic Energy Commission bef'ore resumption of'nit operation. Operation beyond such a limit may not fn itself result fn serious consequences but it indicates aa operations'eficiency subject to regulatory reviev.B.-The limiting safety system settings are settings on fnstramcntatfon vhfch iaftfate the automatfc protectfve action at a level such that the safety limits vill not be exceeded.Thc region'bctveen thc safety limit and these settfngs represents margin vfth normal operation lyfag be)ov these~ettings.Thc margin has been established so that vith proper operation of the instrumcntatioa the safety limits vill never.be exceeded.C-The limftfag conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility.%hen these conditions are met, the plant can be operated safely and abnormal sit t ollcd g~/ac,s.WB l 4 propos M.l-.C O 3.c.3 Zn thc cat a L Coadft for erati an or sociate equircmeat caaaot be tis d bec e o ci taace fn excess o those ad ssc the spec cation, e unit s be place fa a least ot taadby thin 6 urs and in ld Shut vi in th loving 0 hours ess correc ve mess s ar co eted t permit eration the pe issi e isco or til the r tor is pla d ia aa era onal c itio in the spec ication is ot appl able.Exc tions thes requircm s shall be tated the i al sp ifica ons.Thi rovides a fons to c t en fo cir tance ot direct provided or in spc ficat ere currence uld viola the faten of the peci atioa.For exam e, if a ciffc ion c s fo tvo terna ('or bsystems)o be opc ble prov s for li t rcquir ts if on syst (or s system s in crab then i oth syst r sub tcms arc ino able the t is to e in at lc t Hot andb in 6 ho an n Co Shutdo vithfn the lovi 0 h rs if th inop able ondition is not orrect Aoo Abb APD ADD ADD BFB Unit 2~w l CO L Co@CO Z,o.1 p.O.4'3 eOB 9 P,OB 5 3',o 5 Lc.O 3 0.7 Bases 1.0-1 i~FV~L".2 o~ 1.0 See JV5)s'fiC4li4+ Z4r C4CeJ+gP~/S~S],o (Cont'd)S'>'l>C 4'On Ps 773 owned, leased,, or otherwise controlled by TVA.-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 industrial. commercial, institutional, or recreatiohal purposes.IZ.-The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in CcCi/gm)which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, Z-134, and'-135 actually present.The thyroid dose convezsion factor used for this calculation shall be those listed in Table IZI of TZD>>14844"Calculation of Distance Factors for Power and Test Reactor Sites".-The chazcoal adsorber vessels installed on the discharge of the steam get air ejector to provide delay to a unit's offgas activity prior to release.-An individual in a controlled or UNRESTRICTED AREA.However, an individual is not a MBMBER OP THE PUBLIC during any period in which the individual receives an occupational dose (as de ed in 10 CPR 20).g'(ala m 4gQ Pr4 C4 R X.o.p<p/~t I~<P~~,5'R 3.Ct.2.ps Pf'I L uzvel cgllze s met ing the OPERAT NAL CONDITZONS or oth conditions specified fo individua limiting ditions for operatio unless othe&indivi ts.Bach Surve lzemen s per 0 w elf ied suzveillanc interval with a maxi allowable extension ot to exceed 25 percent f the specified surveil ance interval.Zt is t intended that this (tension)provision used repeatedly as a venience to extend surveillance intervals nd that specified for surveillances that are no erfozmed durin refueling outages Performance of a Surveillance Requirement with the specified time into sha11 constitute c iance and OPERABILITY pequirements Eoz a limi ing condition'or operat on and associated action statements unless therwise required by the e specifications. Surveillance remsnts do not have to be erformed on inoperable equipment. 5@7~/Zf it is discovezed that a surveillance was not performed within its specified!renneney, then templienee with the rehnirement tn 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 fzequency, whichever is less.This delay period is permitted to allow performance of the suzveillance. BPN Vnit 2 Add SZ S.o.f P,Ct SC l&zeg 1.0-11 PAGE~OF~ 0' A@3 0,3 If the surveillance is not performed within the delay period, the LCO ((~4.)must iaN(ediately 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 cntcrcd.-Surveillance Requirements for Inservice Testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: Inservice testing of ASME Code Class 1, 2, and 3 pur~s and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Conmission.pursuant to 10 CFR 50, Section 50.55(g)(6)(i).Surveillance intervals specified in Section XZ of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications: ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda'or performing inservice terminology for inservice Sm SuskT~~(<<go~c (t~Pf 6 ((,PV (S75 5.5 Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once.per 92 days At least once per 184 days At least once per 276 days At least once per 366 days The provisions of Specification 1.0.LL are applicable to the abc'equired frequencies for performing inservice testing activities. Performance of the above inservice testing activities shall be in addition to other specified surveillance requirements. 5.Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification. 6.The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with.".c staff positions on schedule, methods, personnel, and sample expansion included in this generic letter.BFN Unit 2 1.0-12 S UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP

I 0 Sec r~s~cat'on ~Cha yes W BP<lS f5 5Cch'on SPec.';c~4on s.o The aucceediag frequently used teraa are eXplicitly defined ao that a md,fora interpretaticm of the speci ficationa nay be Lchiovedo sealdala-reasonable aafntenance of the claddiac Lnd prfaary ayateas are assured.Ezceeding such a liaft requfzea aait ahatdovn Lnd rerfev by the Suelear Reulatory Cceaiaafon before reatmytfon of aait oyeration. Operatfcm beyond such a lfaft~not fn.itself reaalt fn serious consequences bat it indlcatea an operational deficiency aab]ect to regulatory reriev.B.C.-The lbd.thg safety ayatea setting+Lre settinas on fnstzl&entaticm vhfch initiate the LntoRatfc protectfve action at a level such that the safety lialts vill not be exceeded.Tha refon betveen the safety lait and these settings reyreaenta aargfn vlth noraal operation lying belov these settings.The aaron haa been established ao that vith yroyer operation of the fnstnmeatatfcm the safety 1fjifts vill never be exceeded.-The liaitiag conditions for oyeratlcm syecity the sigma acceytable levels of syatea perforaance necessary to assure safe atartap snd oyeration of the facility.Whea these conditions are aet, the ylant can be oyerated.safely snd abnoael aitaatfons can be aafelg controlled. Rfg)a~~;+k PI octa g~Cco 3y g Ql ths ovent 4 t tion and/0 LssocfL ed reqsfrenents cannot.be sLtisfied beca0$of cf T tances fn<<xcess of tho 4ddTesaed fn the speci cationy unit shall be yla in at less Hot S vithin 6 snd fn Cold tdovn vi the foll 30 ars aaleaa orrectfve ae ea are c lated that craft oyeratf under the y aalble di Tery or an il the reactor is laced in an perational tfcm fn ch e yeciffcati ia not ayyl cable.Exc thms to ae r reaenta 1 be stat in the fnd vfdaa1 spe f feat f ons o s proTi action to e tLkoh f 0 cir tancea not frectly y ded for the speclff tlcms Lnd ere oc e voald olate" the" intent of ayecf fi tion.2'o exaaple, i a elf lcatf calls fo tvo ayst (or aab eas)to be o rabie and Topsides fo explicit equi reaenta f cme sya (or tea)is yerable,<<n if bo systems (or teaa)are inoyera le the anl ia to be at least ot St fn 6 hours in Col Shatdovn thin the fo 30 if the oyerable ondition ia not coTTect N>boo Lco Na uo Lco~co3~oo ceo~c.o.-neo c cco (Q~-Roo Bm AM Ado Unit 3 rr3 (gl goo LQ Zo.I 3i0.Q 3.0, 9 g.o,S P.o.b 3o0e 7~es.1.0-1 PAGE~QF~

1.0 ScC WuSHf&$lon g~ch~nys h 8pn1 gsv5 Sec~on).0 (Con'd)sPcci+c'~tjon 3~o 15 373 owned, leased, or otherwise controlled by TVA.-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 industrial, commercial, institutional, or recreational purposes.ZZ.-The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in pCi/gm)which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, Z-132, Z-133, Z-134, and I-135 actually present.The thyroid dose conversion factor used for this calculation shall be those listed in Table Z1Z of TZD-14844"Calculation of Distance Factors for Power and Test Reactor Sites".-The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.-An individual in a controlled or UNRESTRICTED AREA.However, an individual is not a MEMBER OF THE PUBLIC during any period in which the individual receives an occupational dose (as defined in 10 CFR 20).QPlacf eifh Prof O3;~s'R'.o.(-Surveillan e Requ'rements hall b met uring>the OPE TZONAL ONDIT ONS or ther c ndition specif ed fo indigidual limit g con tions or ope tion le e in indivi ual Survei e Re ac urve3.ance equxrement e performe with n t e sp cx@ed s eillance interva with a maxi all able exte sion not to exceed 5 perce of the pecified surve lian interval It is ot intend d that t is (exte ion)prov'ion e used r eatedly as a conv ience to extend su eillanc intervals beyond t t speci ied for surveillance ed durin refueling utage~~PJrue e+h P p~~Sg 3.s.Z.Performance of a Surveillance Requirement within the spe 3.me interval shall constitute comply ce and OPERABILITY requirements for a limitin condition for operatio and associatedQction statements unless othe ise required by these pecifications. Surveillance irements do not have to be erformed on ino e i ment.t3 Zf it is discovered that a Surveillance was not performed within 3.ts gg+P specified frequency, the compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to hours or up to the limit of the specified frequency, whichever's less.This delay period is permitted to allow performance of the Surveillance. BFN Unit 3 Add 5R 3.o.+Q aid SR~g 1.0-11 pp, i~at'~ 0 Sgf.0.3 Zf the surveillance is not performed within the delay period, the LCO (co4.)must irrgnediately be declared not met, and the applicable condition(s) must be catered.When the surveillance is performed within the delay period and the surveillance is not mct, the LCO must immediately be declared not met, and the applicable condition(s) must be entered.Suzve llance Requirements or ec son Z Pump an a ve Program-Surveillaace Requirements for Znservice Testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: Znservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XZ of the ASME Boiler and Pzessure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission puzsuant to 10 CPR 50, Section 50.55a(g)(6)(i).2.Surveillance intervals specified in Section XZ of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these technical specifications: ~ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda'or performing inservice terminology for inservice 8C Smfifi'ek>< foe C44~+>9&8pw IS.rS CS Meekly Moathly rterly or every 3 months emiannually or every 6 months Every 9 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184.days At least once per 276 days At least once per 366 days 3~The provisioas of Specification 1.0.LL are applicable to the above required frequencies for performing insezvice testing activities. Performance of the above inservice testing activities shall be ia addition to other specified surveillance requirements. 5.Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the rcquiremcnts of any technical specification. 6.The inservice inspection program for piping identifcd in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods, personnel, and sample expansion included in this cneric letter.BFH Unit 3 1.0-12 JUSTIFICATION FOR CHANGES SECTION 3.0-LCO APPLICABILITY ADNINISTRATIVE CHANGESA1 A2 The Bases for Section 3.0 are being added in accordance with applicable Bases from NUREG-1433. The individual changes made to add LCO 3.0 and SR 3.0 Technical Specification provisions provide the justifications necessary to substantiate the contents of these Bases.LCO 3.0.6 contains new provisions over present Technical Specification requirements. These new provisions provide guidance regarding the appropriate actions to be taken when a single inoperability (e.g., a support system)also results in the inoperability of one or more related systems (e.g., supported system(s)). In the existing Technical Specifications, along with their intent and interpretation provid~d by the NRC over the years, there is not an unambiguous approach to the combined support/supported system inoperability. Guidance provided in the June 13, 1979, NRC memorandum from Brian K.Grimes (Assistant Director for Field Coordination) would indicate an intent/interpretation consistent with the proposed LCO 3.0.6-without the necessity of also requiring the additional actions of a Safety Function Determination Program.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 the impact of a support system on the remainder of the Technical Specifications, would indicate a similar philosophy of not taking actions for the inoperable supported equipment. However, in this case, additional actions similar to the proposed Safety Function Determination Program actions, were addressed and required.Generic Letter 91-18 and a plain-English reading of the existing STS provide an interpretation that failure to perform a required function, even as a result of a Technical Specification support system, requires all associated action be taken.Considering the history of disagreement and misunderstandings in this area, the BWR Standard Technical Specification, NUREG-1433, was developed with industry input and approval of the NRC to include LCO 3.0.6.The present custom BFN Technical Specifications do not contain specific Action Statements as presented in current STS plants.The present BFN Technical Specification provisions are usually left to interpretation as to supported system operability when a support system is inoperable unless specific direction is contained in the Technical Specification. Since the addition of LCO 3.0.6 to the BFN Technical Specifications will clarify existing ambiguities and maintain actions within the realm of previous interpretations, this new provision is deemed to be administrative in nature.BFN-UNITS 1, 2,&3 Revision 0 PAGE/

JUSTIFICATION FOR CHANGES SECTION 3.0-LCO APPLICABILITY ADMINISTRATIVE CHANGES (Continued) A3 LCO 3.0.7 is a new requirement that is not presently addressed in the BFN TSs.Present BFN TSs do not contain special operations LCOs and, as such;do not need LCO 3.0.7 allowances. LCO 3.0.7 specifies that compliance with special operations LCOs is optional, but if used, then the actions of the special operations LCOs shall be met.The applicability of a special operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS.A special operation may be performed either under the provisions of the appropriate special operations LCO or under the other applicable TS requirements. When a special operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCOs applicability. The surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO.This Specification eliminates the confusion which would'therwise exist as to which LCOs apply during the performance of a special test or operation. This change is considered to provide administrative controls for the use of Special Operations LCOs and, as such, this change is administrative in nature.Editorial rewording, reformatting and renumbering is made consistent with the BWR Standard Technical Specifications, NURfG-1433. During the development of SR 3.0.1, certain wording preferences or language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. The second sentence of SR 3.0.1 has been worded to clarify an existing intent that is not explicitly stated.The provision states that failure to meet a Surveillance Requirement can occur during performance of the SR or between performances of the SR and either case shall result in failure to meet the LCO.A5 The last sentence of SR 3.0.1 is clarified by adding the phrase"or variables outside specified limits." This addition is necessary since all LCOs do not deal exclusively with equipment operability. Editorial rewording for SR 3.0.2 is made consistent with the BWR Standard Technical Specifications, NUREG-1433, which has resulted in no technical changes (either actual or interpretational) to the basic application of the 25%extension to routine surveillances. Also, the sentence Exceptions to these requirements are stated in the individual Specifications" is added to acknowledge the explicit use of exceptions in various surveillances. These changes provide consistency of wording and clarity for understanding. No technical changes (either actual or interpretational) to the Technical Specifications are intended by these changes.BFN-UNITS I, 2,&3 Revision 0 rAca>~"~V JUSTIFICATION FOR CHANGES SECTION 3.0-LCO APPLICABILITY TECHNICAL CHANGE-NORE RESTRICTIVE N3 N4 Present BFN Technical Specifications do not contain specific guidance such as proposed in LCO 3.0.1.Present individual Technical Specification requirements are understood to apply as stated in each individual specification without the necessity of a motherhood statement. The adoption of NUREG-1433 for BFN will add requirements not presently in the Technical Specifications and, as such, potentially more restrictive provisions. Proposed LCO 3.0.1 contains exceptions to LCO 3.0.2 and LCO 3.0.7.These exceptions will state when it is acceptable for the LCO to not be met during the Nodes or other specified conditions in the Applicability. The exception to LCO 3.0.2 is necessary since LCO 3.0.2 addresses the condition of meeting the associated Actions when not meeting a LCO.The exception to LCO 3.0.7 is necessary since LCO 3.0.7 provides special guidance that will allow Special Operations Technical, Specifications to govern over the LCOs in Sections 3.1 through 3.9.Present BFN Technical Specifications do not contain the provisions of LCO 3.0.2.LCO 3.0.2 requires that upon failure to meet an LCO, that the required Actions shall be met.Present BFN custom Technical Specifications do not contain a specific Action Statement section.The Actions for the present BFN Technical Specifications are contained within paragraphs that also contain the LCO and Applicability. Exceptions to LCO 3.0.2 are as provided in LCO 3.0.5 and LCO 3.0.6.These exceptions are necessary to allow systems to be returned to service under administrative control to perform the testing required to demonstrate operability per LCO 3.0.5, and to allow a supported system to be considered operable and its Actions not to be taken, solely due to a support system inoperability per LCO 3.0.6.LCO 3.0.2 also contains a provision such that completion of the Required Actions is not required, unless otherwise stated, if the provisions of a LCO are met or are no longer applicable. Since the present BFN TSs do not contain provisions similar to LCO 3.0.2, this proposed change may result in potentially more restrictive requirements. The present BFN TSs do not contain provisions similar to proposed LCO 3.0.4.LCO 3.0.4 prohibits entry into a mode or other specified condition, when a LCO is not met, unless the associated actions to be entered permit continued operation for an unlimited period of time.The provisions of LCO 3.0.4 will 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 will not prevent changes in modes or other specified conditions in the applicability that result from a normal shutdown.Any exceptions to this specification are stated in individual specifications. The addition of this specification to BFN TSs represents a potentially more restrictive change.The sentence"For Frequencies specified as"once," the above interval extension does not apply" is proposed to be added.The interval extension concept is based on scheduling flexibility for repetitive performances, and these"once" surveillances are not repetitive in BFN-UNITS I, 2,&3 Revision,O Wee~<<~ JUSTIFICATION FOR CHANGES SECTION 3.0-LCO APPLICABILITY TECHNICAL CHANGE-NORE RESTRICTIVE (continued) nature and essentially have no interval as measured from the previous performance. The nature of these SRs precludes the ability to extend their performances, and is therefore a more restrictive requirement. The existing specification can be interpreted to allow the extension to apply to all surveillances. H5 Present BFN TSs do not contain provisions similar to proposed 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.This specification applies to changes in modes or other specified conditions in the applicability associated with unit shutdown as well as startup.The provisions of SR 3.0.4 will not prevent changes in modes or other specified conditions in the applicability that are required to comply with actions.TECHNICAL CHANGE-LESS RESTRICTIVE Ll Present Oefinition 1.0.C.1 contains the requirements typically refereed to as LCO 3.0.3 in the Standard Technical Specifications. The changes to present requirements include the following: One'additional hour for plant shutdown is included over present provisions. This additional hour was included in the STS to allow time to initiate a plant shutdown.The additional hour is included in the ISTS to allow initiation of the plant shutdown and is included in the ISTS times allowed to Node 2 of 7 hours, to Hode 3 of 13 hours, and to Hode 4 of 37 hours.Present provisions require the unit to be in Hot Standby within 6 hours and in Cold Shutdown within the following 30 hours.The ISTS requires the unit to be in Node 2 (Startup/Hot Standby)within 7 hours, Hode 3 (Hot Shutdown)within 13 hours and Hode 4 (Cold Shutdown)within 37 hours.This change will relax present shutdown provisions since Hode 2 can be entered at less than approximately 15%rated thermal power whereas present requirements to be in Hot Standby would require the unit to be less than or equal to 1%rated thermal power within 6 hours.This is offset by a more restrictive requirement to be in Hode 3 in 13 hours.Present BFN Technical Specifications do not contain the provisions-of LCO 3.0.5 from NUREG-1433. LCO 3.0.5 is added to provide an exception to LCO 3.0.2 for instances where restoration of inoperable equipment to an operable status could not be performed while continuing to comply with required actions.Hany Technical Specifications actions require an BFN-UNITS 1, 2, 8i 3 Revision 0 PAGE QF

JUSTIFICATION FOR CHANGES SECTION 3.0-LCO APPLICABILITY TECHNICAL CHANGE-LESS RESTRICTIVE {continued) inoperable component to be removed from service, such as maintaining an isolation valve closed, disarming a control rod, or tripping an inoperable instrument channel.To allow performance of Surveillance Requirements to demonstrate the operability of the equipment being returned to service, or to demonstrate the operability of other equipment which otherwise could not be performed without returning the equipment to service, an exception to these required actions is necessary. LCO 3.0.5 is necessary to establish an allowance that, although informally utilized in restoration of inoperable equipment, is not formally recognized in the present Technical Specifications. Mithout this allowance certain components could not be restored to operable status and a plant shutdown would ensue.Clearly, this is not the intent or desire that the Technical Specifications preclude the return to service of a suspected operable component to confirm its operability. This allowance is deemed to represent a more stable, safe operation than requiring a plant shutdown to complete the restoration and confirmatory testing.L3 The sentence"If a Completion Time requires periodic performance on a'once per...'asis, the above Frequency extension applies to each performance after the initial performance" is proposed to b'e added.This provides the consistency in scheduling flexibility for all performances of periodic requirements, whether they are surveillances or required actions.The intent remains to perform the activity, on the average, once during each specified interval.BFN-UNITS I, 2, 5.3 Revision 0 PAG~~~o"~~~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP

3.3 4.3 hpplies o the opera onal status o the control od system.kppllea t the surveillance requireaen of the control rod systea.To assure the billty of th control rod sya ea to contro reactivity. To verify the a illty of the control rod'syst to control reactivity. S pA/3>I.I SO@i SPA>o 1~f L sufficient number of con-trol rods shall be operable so that the core could be sade subcritlcal in the sost reactive condition during, the operating cycle vith the strongest control rod fully vithdravn and all other operable control rod fully inserted.SR 3.l.l.l.a Pn posed g"~~us L.41 S fici t co+rog 1 be kiehkra 0'uc outage vhen core alterations ver erf 0 r e a margin of 0.38K h k/k the core can be made aubcritical at any time in the subsequent fuel cycle vith the analytically determined stroxgest operable control rod fully vlthdravn and all other operable rods full inserted.~R 3.1.I,l i 5 7'c'quent~ BFH Unit 1 3.3/4'-1

~~'t~\k~~~~~~~~~~~~~~~.~II~~'~~~~~~~~~~~~~~~~~'~~~~~~~~'~~~.-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~t~~~~~~~~~~~~~~~~0~~~~~~~~~~~' UNIT 2 CURRENT TECHNICAL SPECIFICATION IVIARKUP 0

3.1 Sysv'6MS

LINITIHG COHDITIOHS POR OPERATIOH 3.3 REACTIVITY COHT1%L A lica ilit SURVEILLANCE RBQUIREHENTS 4~3 REACTIVITY COHTROL Applies to the operational status of t control rod s tea.Applies to the rveillan e equirements of t c contre system+PA, o~b eetivc T assure the abilit of the con ol rod system to trol react ity.SDQ 3el.I Shufdo~n (snnn)S.l.t Mar in Lear~~Ob tive To ver y the ability the control od system tn c trol reactlvit S ccification L.Cd 3 I.l.~V-.p..@L.c5 ZI,I J, A sufficient number of con-trol rods shall be OPERABLE so that the core could be made subcritical in the most reactive condition during thc operating cycle with the strongest control rod fully withdrawn and all other OPERABLE control rods ully inserted.Ml Acr<oh/$p 8 g y+g sR 3.I.I.I Pd pt f4'ry+t S flcle t control rods hall vlt@ravn o loving a refueling outage when core alterations vere crforaed o e s rate vith a margin of 0.38%4 k/k the core can be made subcritlcal at any time in the subsequent fuel cycle with the analytically determined strongest OPERABLE control rod fully withdrawn and all other OPERABLE rods fully inserted.rsvp'rppuc~cp 7rdtdfa4 SX R./I.I.X BPH Unit 2 3.3/1.3-1 PAGE 0 0 5 ec'Ac k'~~~.l I APR 3 0 ssss ont d.DELETED e.Control rods with inoperable accumulators or those whose position cannot be positively determined shall be considered inoperable. g.ed Z.I.noperable control rods shall be positioned such that Specification 3.3.A.l is met.In d.The control rod accumulators shall be determined OPERABLE at least once per 7 days by verifying that the pressure and level detectors are not in the alarmed condition.

• r RFtv lsrz kl,5 Pwiog B addit on, during reactor power operation, no more than one control rod in any Sx5 array may be inoperable (at least 4 OPERABLE control rods must separate any inoperable o If this specification cannot be met the reactor shall not be started, or if at power, the reactor shall be brought to a S OWH COHDITIOH within ours.l?~ce.~iskliccfio~

gyt Qt r,~+~FPN is~g ig BFH Unit 2 3.3/4.3-4 AMENDMERT HO.2 g 2 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 0 Z.j 3 3 4.3 kppl s to the op ational stat of the cont 1 rod system.Applies o the survei ance requirem ts of the co rol rod system To assure ability of he control rod stem to cont ol reactivityo To verify the a lity of the con ol rod syst to control reac vity.Shu).boa>3,1,1 331~)/SPY)50+3)~(fbo~osed LCp pj f b k sufficient number of con-trol rods shall be OPERhBLE so that the core could be made subcritical in the most reactive condition during the operating cycle vith the strongest control rod fully vithdravn and all other OPERABLE ccmtrol rods fully inserted.5'R3.l,jf g S fic c tr ro sha'kl fo loving a refueling P~<outage vhen core A~~(~ct alterations vere performed to emonstrate vith a margin of 0.38K h k/k the core can be made subcritical at any time in the subsequent fuel cycle vith the analytically determined strongest OPERABLE control rod fully vithdravn and all other OPERkBIS rods fully inserted.~R 3<~I t]H%Qpc'n<y~~RC-Tiong 4 6 Q O gF: BFH Unit 3 3.3/4.3-1 PAGE d.DEIZXED e.Control rods vith inoperable accumulators or those vhose position cannot be positively determined shall be considered inoperable. The control rod accumulators shall be determined OPERhBLE at least once per 7 days by verifying that the pressure and level detectors are not in the alarmed condition. Wood 8 Inoperable control rods shall be positioned such that Specification 3.3.h.l is met.In addition, urine reactor pover operation, no more than one control rod in any 5x5 array may be inoperable (at least 4 OPERMKZ control rods must separate any inoperable ones).If th s spec cat on canno't be met the reactor shall not be started, or if at pover, the reactor shall be brought to a S MH COHDITIOH vithin hours.>~3 QS54'(~op Qr chal~+~8<<asks z.I.Z~854%cA50~ $r gQ&BF~~STS p-l.3 BFH Unit 3 3.3/4.3~AMENDMENT ll0.I 6 9 3,~-...3 0 'USTIFICATION FOR CHANGES BFN ISTS 3.1.1-SHUTDOWN MARGIN (SDN)ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 The LCO has been reworded to include the actual limit.CTS describes how to demonstrate conformance to the limit, however the actual limit is located in the corresponding surveillance requirement. The required limit when the highest worth control rod is analytically determined (0.38%~k/k)is included for clarity.A3 The proposed Surveillance Requirement provides a specific completion time to clarify when the SDH verification is to be completed. The intent of present Technical Specification 4.3.A.1 is to require the SDN test to be performed after in-vessel activities which could have altered SDN.Nore explicit wording is proposed to replace the activity referred to as"following a refueling outage when core alterations were performed." Host SDN tests are performed as an in-sequence critical.The proposed Frequency of 4 hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification. This interpretation is supported by the BWR Standard Technical Specifications, NUREG-1433. Since the proposed change clarifies the intent of the existing Surveillance Requirement, it is considered an administrative change.Both limits described in Comment Ll below are also listed in the Surveillance. Thus, the limit for the highest worth rod determined by test (0.28%a,k/k)has been added.TECHNICAL CHANGE-NORE RESTRICTIVE Currently, if SDM is not met the unit is placed in a Shutdown Condition (Mode 3)within 24 hours per CTS 3.3.A.2.f. Proposed Action B requires the plant to be placed in Mode 3 if SDH is not met.Proposed Actions C, D, and E for Nodes 3, 4, and 5, are more restrictive than CTS since some additional action is required if SDN is not met (e.g,, insert all insertable rods, suspend core alterations, initiate action to restore secondary containment to OPERABLE status, restore two standby gas BFN-UNITS 1, 2, 5.3 PAGE OF O evision

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.1-SHUTDOWN MARGIN (SDM)TECHNICAL CHANGE-MORE RESTRICTIVE CONTINUED treatment subsystems to OPERABLE status and restore one isolation valve and associated instrumentation to OPERABLE status in each secondary containment penetration flow path with isolation valve(s)not i".olated within 1 hour).The following changes were made to current Technical Specifications: ~If SDN is not met while the plant is in Node 1 or 2, the proposed Actions (A and 8)would require SDN to be restored in 6 hours or be in.Node 3 in the following 12 hours.Therefore, the proposed Specifications are more restrictive since only 18 hours is allowed to be in Node 3.In addition, once in Node 3, if the SDN was still not met, Action C would require the insertion of all insertable control rods.This action further enhances the available SDN.Since the plant was shut down to get to NODE 3, then the only action required is to insert all insertable control rods since secondary containment, standby gas treatment and isolation instrumentation are all required to be operable in NODE 3 anyway.~If SDH is not met in NODE 4 or 5, new ACTIONS (ACTIONS 0 and E)are provided to initiate action to insert all insertable control rods (in core cells containing fuel), suspend CORE ALTERATIONS (if applicable), and to initiate actions within 1 hour to restore secondary containment, SGT System and the secondary containment isolation valves to OPERABLE status.The first two actions attempt to improve SDH, or at least to ensure SDH is not made worse, while the last three actions provide some protection from radioactive release if a SDN problem results in an inadvertent criticality. These Actions are more restrictive since new requirements are added that currently do not exists.H2 An additional Surveillance Frequency for SDN verification (prior to each in-vessel fuel movement during fuel loading sequence)has been added to clarify the requirements necessary for assuring SDH during the refueling process.Because SDN is assumed in several refueling mode analyses in the FSAR, some measures must be taken to ensure the intermediate fuel loading patterns during refueling have adequate SDN.This change imposes a requirement where none is explicitly provided in the existing Technical Specifications. This new requirement does not, however, BFN-UNITS 1, 2,&3 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.1-SHUTDOWN MARGIN (SDM)TECHNICAL CHANGE-MORE RESTRICTIVE CONTINUED require introducing tests or modes of operation of a new or different nature than currently exist.As presented in the Bases corresponding to this requirement, this is best accomplished by analysis (rather than in-sequence criticals) because of the many changes in the core loading during a typical refueling. Bounding analyses may be used to demonstrate adequate SDM for the most reactive configurations during refueling thereby showing acceptabili.ty of the entire fuel movement sequence.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Details of the methods to perform the Surveillance are relocated to the procedures. The requirement to verify the SDM is within the limit remains in the Surveillance. Procedures will be controlled by the licensee controlled programs."Specific" Ll The current Technical Specifications indirectly require that the SDM be a 0.38~k/k when the highest worth control rod is analytically determined. NUREG-1433 adds a less restrictive requirement that allows the SDM to be a 0.28 u,k/k when the highest worth control rod is determined by test.This allows the SDM to be less when the highest worth control rod is determined by test.This is reasonable since the highest worth control rod is directly calculated. An actual measured value is obtained for the highest worth control rod versus an analytical one which may contain uncertainties that have to be accounted for in the analysis.The Surveillance Requirement also incorporates the new SDM value.BFN-UNITS 1, 2,&3 pAGE~O.Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP SPecikmh'on 3.l.2, GEC 0 7 1994 I-ACT(oN If Specifications 3.3.C and.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the S UTDOWN CONDITION within hours'2 Su~eill nce reqyiremegts are as Qecif ed in 4 3.C ahd.D above The scram discharge, volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.Z.l.a.The scram discharge volume drain and vent valves shall be verifie open PRIOR TO STARTUP and monthly thereafter. The valves may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation l.b The scram discharge volume drain and vent valves shall be dcmonstratcd OPERABLE in accordance with Specification 1.0.MM.2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided thc rcdundant-drain or vent valve is OPERABLE.2~When it is determined that any SDV drain or vent valve is inoperablc, the redundant drain or vent valve shall bc demonstrated OPERABLE immediately and~eekly thereafter. 3.If redundant drain or vent valves become inoperable, the reactor shall bc in HOT STANDBY CONDITION within 24 hours.3~No additional surveillance required.BFN Unit 1 3.3/4.3-12 AMEN%SrlI.P g>

SP<c.itic<on 3 t 2.3.3.C.4.3.C.X Inserted From Avg.Scram Inser-t se 5 20 50'0 0.398 0.954 2.120 3.800 3.The maximum scram insertion time for 90K insertion of any operable control rod shall not exceed 7.00 seconds.2.The average of the scram inser-tion times for the thrcc fastest operable control rods of all groups of four control rods in a tvo-by-tva array shall be no greater than: 2~At 16-vcek intervals, 10K of the operable control rod drives shall bc scram-timcd abave 800 psig.Whenever such scram time measurements are made, an evaluation shall be made to provide rcasonablc assurance that proper control rod drive performance is being maintained 5'e~guStif'act,h'e~ Par Change A BRA}ST5 3,}.Q/Vlcc.os.confro}.fed glee,ewe~S p".Ll'.o P.l.2~RogS A+8 The reactivity equivalent of the diffcrencc bctveen the actual critical rod configuration and thc expected configuratio e ll not cxcced 1X hk If this limit is exceeded the r actor vill be 1 t 1 thc ause s e et rm ned~d c rrc ivc ac ions vc ecn t~cn as app opri e.f'rcqosW R<q~'red 4Hon A.}L}M3,}>l Duri e st t s 0 r 0 av uc ng out cs e critical rod configurat ons vill be compared to the expected configurations at selected operati conditions. Th sc corn risons vi e us ba dat fo rea tiv ty ito ng during ubs qu t po er o ra ion hro au e fuel cle At specific povcr operating conditions, the critical rod configuration vill be compared to the configuration expect e by,scd$upap p<o riathl cbgrdetc past a~.This comparison bc made at least eve'all paver month.BFH Uni.t 1 3.3/4.3-11 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

3.3.C.o es 4.3.C.S s o es 2.The average of the scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a tvo-by-tvo array shall be no greater'than: Z Inserted From Avg.Scram Inser-5 20 50 90 0.398 0.954 2.120 3.800 t 3.The maximum scram insertion time for 90Z insertion of any OPERABLE control rod shall not exceed 7.00 seconds 2.At 16-meek intervals, 10Z of the OPERABLE control rod drives shall be scram-timed above 800 psig.Whenever such scram time measurements are made, an evaluation shall be made to provide reasonable assurance that proper control rod drive performance is being maintained. ~<<<<<~'<'mba A~CS~>+>+BF/V lg~~~~y jL 1 5R 3.l.z.(Ae$s A+8 The reactivity equivalent of the difference betveen the actual critical rod configuration and the expected configuratio s all not exceed IZ 4k If this limit is exceeded t e r be lace ghe us as en+term ed+d o ect ve a tio+have bee ake as pro riate During t e and AR o oven e outa es the critical rod configurations vill be compared to the rCp4~~<>t expected configurations at selected operating conditions. ese omparisons v 1 be ed a base ta for r ctivi moni oring urin sub quent pover perat n throu out the fue c c At spec fic pover operating conditions, the critical rod configuration vill be compared to the configuration expected a ed upon p r keel cor%ecteM a a This comparison vill b~made at least every>full<oozier month.BFN Unit 2 3.3/4.3-11 AMENDMENT NO.'yp'5 0 0 j j-p,c Ziod 8 Hl l If Specifications 3.3.C and.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the UTDOMN CONDITION within hours.Su eillance uirements are specified'n 4..C and.D a 4.3.F.1.The scram discharge volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.l.a.The scram discharge.volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter. The valves may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation. 2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.l.b.2.The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MN.When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE iamediately and weekly thereafter. 3-If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBY CONDITION within 24 hours.3~No additional surveillance required.S<-4<f~~io4fio~ 44t-Ch<~pg<W l~S g./q BFN Unit 2 3.3/4.3-12 NEemrr Ng.g 29 PAGE~OF 3 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 0 NOV ie Zae 3.3'D 4.3.C.S se 0 2.The average of the scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a tvo-by-tvo array shall be no greater than: Z Inserted From Avg.Scram Inser-5 20 50 90 0.398 0.954 2.120 3.800~i hludc'S I 3.The maximum scram insertion.time for 90K insertion of any OPERABLE control rod shall not ceed 7.00 seconds.The reactivity equivalent of the difference betveen the actual critical rod configuration and the ected configuration s 11 not exceed 1X hk.Pg,pseud Pew,~Acm~/t I If s limit is exceede th++8 4 reactor vill be pla ed I e s ermin d c rrect ve a iona ha e b en t4cen a appro riate 2.At 16-week intervals, 10K of the OPERABLE control rod drives shall be scram-timed above 800 psig.Whenever such scram time measurements are made, an evaluation shall be made to provide reasonable assurance that proper control rod drive performance is being maintained. ~<>~SHF'~b'~A 4 C Chavez Swv l575 3 l 9 Oa Co~+.oL A J g)ace~A+SR3.1.2 I (luring ta st and art o 0%K Ilg out es the critics rod configurat ons vill be compared to the expected configurations at selected operating conditions. ese omp r s d bas da a fo re tivi m nit ring du sub quen po er pera io thro out the fue c e.At specific pover operating, conditions, the critical rod configuration vill be compared to the confi uration cted a e u n pp o+i tel co e e t&R~This comparison vill be made at least every;uLJ sawer month..BFK Unit 3 3.3/4.3-11 AMENDMBtT NQ.g p g PAGE~OF~ 0 0 0 7 1994 g.%ion 6 3-3.F.If Specifications 3.3.C and 3.3.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall bc in the S OWN ONDITIO within hours./2.4.3.F.S are and eill ce rcqui emcqts as spe ified in 4.+C.3.D a ve.1.The scram discharge volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.l.a.The scram discharge volume dra.in and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter. The valves may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation. l.b~The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance <<ith Specification 1.0.MM.2.In thc event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.2~When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE iamediately and<<eckly thercaftcr. 3.If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBY CONDITION within 24 hours.3~No additional surveillance required.BFN Unit 3 3.3/4.3-12 ~3AShf-iat6an Qc Cj4~gg kr a~rJ l57$p.l.8 AMBlDhlENT gD, y 8 6 PAGE~GF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.1.2-REACTIVITY ANOMALIES ADMINISTRATIVE CHANGES Al Reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS)should be more readily readable, and therefore understandable, by plant operators as well as other users.During the reformatting and renumbering of the improved Technical Specifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. A2 Proposed BFN ISTS LCO 3.0.4 does not permit entry into MODES unless the associated ACTIONS to be entered permit unlimited continued operation. The proposed Specification does not permit exit from MODE 3 (or entry.into Mode 1 or 2)until the reactivity difference is restored.This is considered equivalent to the CTS wording of"until the cause has been determined and corrective actions have been taken as appropriate." Therefore, deleting these words are considered administrative. A3~A<Deleted"During the STARTUP test program" since this event has occurred and cannot occur again.Proposed SR 3.1.2.1 provides a specific completion time for the reactivity anomaly surveillance to clarify when"during each startup" the test must be performed. The test is performed by comparing the actual rod configuration to the vendor provided predicted rod configuration as a function of cycle exposure while at steady state reactor power condition. A time frame of 24 hours after reaching these conditions is considered reasonable to allow performance of the required calculations for verification. This interpretation of the intent of the existing requirement is supported by the BWR Standard Technical Specification, NUREG-1433. Therefore, the proposed change is considered administrative. TECHNICAL CHANGES-MORE RESTRICTIVE CTS require the unit to be placed in the SHUTDOWN CONDITION (reactor in shutdown or refuel mode)if the specified limit is exceeded.CTS 3.3.E requires an orderly shutdown to be initiated and the reactor be placed in the SHUTDOWN CONDITION within 24 hours.Proposed BFN ISTS is more restrictive since it requires the unit to be placed in MODE 3 (Hot Shutdown)within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach MODE 3 from full power BFN-UNITS 1, 2, L 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.2-REACTIVITY ANOMALIES TECHNICAL CHANGES-MORE RESTRICTIVE CONTINUED conditions in an orderly manner and without challenging plant systems.Therefore, the proposed change is considered acceptable. M2 An additional requirement has been added to perform the Surveillance if'ontrol rods have been replaced, regardless of whether or not the unit is in a refueling outage.This ensures that any core change that could affect reactivity is evaluated properly.The Applicability of the Reactivity Anomaly Specification has been expanded from during"power operation" to"MODES 1 and 2." This change represents an additional restriction on plant operations necessary to achieve consistency with safety analysis assumptions and NUREG-1433. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic"~LA1 Details of the methods to perform and purposes of the Surveillance are relocated to the Bases and procedures. The requirement, to verify the reactivity anomaly is within the limit, remains in the Surveillance. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee control programs."Specific" Ll Proposed Action A.1 provides a 72 hour time period to allow the core reactivity difference to be restored to within limits (i.e., to"perform an analysis to determine and explain the cause of the reactivity difference"). Typically, a reactivity anomaly would be indicative of incorrect analysis inputs or assumptions of fuel reactivity used in the analysis.A determination and explanation of the cause of the anomaly would normally involve an offsite fuel analysis and the fuel vendor.Contacting the vendor and obtaining the necessary input may require a time period much longer than one shift (particularly on weekends and holidays). Since shutdown margin has typically been demonstrated by test prior to reaching the conditions at which this surveillance is performed, the safety impact of the extended time for evaluation is negligible. Given these considerations, the BWR Standard Technical Specification, NUREG-1433 allows this time to be extended to 72 hours.BFN-UNITS 1, 2,&3 2 Revision 0 PAGE~OF~ 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION IVIARKUP

5t CCificnbon 5, l~~~Q erg~)AC Or~pi W+g hu io]v+41 gcRIA sec 4l A<Viou]g PPERABip ivy Po~I f'Contro ro vcs ch can not bc moved vith control rod drive prcssure Chal consid n able If a partially or fully vith-dram control rod drive can-not be slovcd vith drive or cram prcssure c react shall bc brought to th SHOTMM?f CO IT vithin hoar and shall not bc started unlcs VCS ation em t ate tha c caas th fa 1 is n fai e con rol r dr me c 11 o and (2)adequate sha ovn margin has been demonstrated as required by Specificaticm 4.3.i.2.c. 4.3.k.2 SR v.).g.2.a.S it'3.l.g.g 43 f&L4red o4 Ae3 Bach partially or fally vi,thdravn OPBRkBLE control rod shall be e c sc onc notch at least once each vc vhen operating above the pover level cutoff of the RW.In thr event pover operation is cmtinu three or more inoperable contro rods s tcs't s be performed at least ncc ea day, vhen operating above the pover lcvcl cutoff of the ROC.Ahfhia 7P Po<pg Pf poi4 Ikbon g,(BFK Unit 1 3.3/4.3-2 AMENDMgtr ga.y 96'an-" 2. 0 ~~~~~r'tt1'lt k+s Ittt)'IVVtII /at)'>>: tibia~'~~t~~~~~~~~\j~~~~~~~~~~~~~~~~~~~arzsa=a CZl Cb~t~~~~~~~~~~4t~~~.t~~~~~A~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~t~~~~~~~~~~~~~~~~~~~~~~~~.~t~~~~~~~~~~4~~~~~~~t~~~~\~t~~t~~~~~t,~~~~~t'~~~~~~~~t~~.~~~~'I~~~~~~

APR 30 Sg 6PEPA85]+/4.3.A.2 (Cont'd)ee se~.t.z.>c~A e 0 f,2.y 4.~)Chandi'on 9 LE Wd't'n E Control rods v th inoperabl cumulatars or ose vhose position cannot be positively determined shall be considered inoperab e.Aj vcl QQ~i5 Inoperable control rods shall be positioned such that Specification 3.3.k.l is met.In a tion u r a or over o eration, no more than one contro rod in any Sz5 array may be.inoperable (at least 4 OPERABLE control rods must separate Lily 2 inoperable ones).If spec on cannot be me re)o sta ted'r at pover the reactor shall e o tto MK COHDITIOH vithin hours.d.e control rod accumulators shall be determined OPERABLE at least once per 7 days b verifying that the pressure and level detectors are not in the alarmed condition. Sc~Zu5ti4c~on 4<CQngc5,+~PfoP~~S ec4i'call~ 3,f.5 fropoSrd Afore W Cobol'Kin J)pq Lco p.o.y aQ~"" P~yp P(opsy Cond,'h'on p~<>ps~A C~gp<cd',Ron s hl+D.2.BFH Unit 1 3,3/4.3~AMENDMENT HD.I 9 6 1'AGE~OF (p <<<<Pc@APR 30593 l.Each control rod shell be coupled to ita drive or c letel naerted and the Pcg)'onI control rod directional control valves disarmed LQ(r s r es ot a pl in e S WSC ZT 5 the r actor i ven ed.ontrol od dri es be r ved long a pecif cation.3.A.l a net~2.The control rod drive housing support ayatea shall be in place during REACTOR ACR OPERATION or Men the reactor coolant ayateE is pressurized above atmospheric pressure vith fuel in the'eactor vessel, unleaa all control rods are fully inserted and Specification 3&.A.l ia net.cv sfgin~rg Verify that the ontrol rod s olla th drive obse ing r sponse e nu lear i t R ation ach i'r d is a ed@hen the re ctor ia o rating above the p eaet pover level cutoff of e RMN.~~'~'~b Shen the rod ia fully vithdravn~q the rst tiae a er each refueli outa e or f e 88 intenance o serve t the drive does not go to the overtravel position+2.The control rod drive housing support systea shall be inspected after reasaeably and the results of the inspection recorded.SrC B.l 3S'M>fbi n shall be verified for 3nourS each vithdram control rod aa follows: BP5 Unit 1 3.3/4.3-5 0 NOV 0 3~3~C I Inserted From Avg.Scram Inser-S A>...5 20 50 90 0.398 0.954 2.120 3~800 SR z,l.3.e The maximum scram insertion time for 90K insertion of any operablc control rod shall aot cced 7.00 seconds.2.The average of the scram inser-tion times for the three fastest operable control rode of all groups of four control rods in a tvo-by-tvo array shall be no great'er than: 4'.C.2~At 16-veek intervals, 10K of the operable control rod drives shall be scraa-timed above 800 psig.Whenever such scram time measurements are made, an evaluation shall be made to provide reasoaablc assurance that proper control rod drive performance is beiag maintained. ~cc T~SNpcation <~ngcS%8FP/Ig gS p].y D.D.The reactivity equivalent of the difference betveen the actual critical rod configuration and the expected coaffguratfon duriag povcr operatioa shall not exceed lX 4k.If this limit is exceeded, the reactor vill be ylaced in the SHOTDOWK COHDITIOS until the cause has been determined and corrective actions have been taken as appropriate. Se<7~H fi cah on Q C Qg~~8P'hl l STs 3, l,2.During the startuy test prograa and startup folloving refueliag, outages, thc critical rod conf fgurations vill be compared to the expected configurations Lt selected operating conditions. These comyarisons vill be used as base data for reactivity moaitoring durfng subsequent pover oyeration throughout thc fuel cycle.At specific pover operating conditions, the critical rod configuration vill be compared to the configuration expected based upon appropriately corrected past data.This comparison vill be made at least every full pover month.BHf Unit 1 3.3/4.3-11 >".ENOMENT NO.I 7 g PAGE 4 DF~ 0 UNIT2 CURRENT TECHNICAL SPECIFICATION MARKUP

5 yc, Awlia~'5.I.3 APR 3 0 1993 f 3.Pgngl ag Acorn,~iz$~I X c'ic,4.ck sh&4.op&'-Aeter. ~~/+~Control rod drives which can-not be moved with control p,<Ylang rod drive pressure shall be j4 g.considered ino erable If a part ally or fully with-drawn control rod drive can-not be moved with drive or scram pressure t e reacto b ou the SHUTDOWN COHDITIOH within ours an s all not be I started l)ves i-g tion as dern tra d t at th cau of th failu e i no a fai ed con ol ro dri mec m collet si d{2)adequate shutdown margin has been demonstrated as required by Specification 4.3.A.2.4.3.A.2 SR 9 I~X a.S/2 X~c4zdw]43 Each partially or fully withdrawn OPERABLE control rod shall be ercise one notch at least once each wee when operating above the power level cutoff of the RM8.In the event power operation is continui wi:h hree or more inoperable control ods t s e shall be performed at least nce each day, when opera ng above the power level cutoff of the RMM.u~<: rz So~44o<g~BFH Unit 2 3.3/4.3-2 AMENDMEÃr go.p y g 0 0 APR 3 0 1893 Add Qg~,r~Ae4'og Q, I R;,g~~.b.e control rod direc-tional control valves for inoperable control rods shall be disarmed e tr4caiJ.b.DELETED 4.3.A.2 Reactivit Mar in-Xn-o crab e Control Rods (Cont'd c.Co rol rod vith sc~~~~times reater an thos permitte by Spe fication.3.C.3 are inopera le, ut t ey can e inserted vith control rod drive pressure they need not be disarmed electrically. Co When it is initially determi d that a control rod is inc able of normal inser on a test shall be conduc d to demonstrate that e cause of thc malfun ion is not a failure in t e con l rod drive meehan m.If this can be demo rated an attempt to ully insert the control d shall be made.If thc ntrol rod cannot be in erted and an investigat n has dern trated that t cause f failure is n t a faile control rod drive me ism collet housing, a utdovn margin test sh ll be made to demonstr te unde this conditi that t core can be made"sub tical for reactivity ndition during the r inder of t operating c e vith the alytically detcrm ed highest vo control r"capable of vithdraval lly vithdravn, an all other control rods capable of insertion fully inserted.BFH Unit 2 3.3/4.3-3~AMENDMENT HP.2 y g PAGE~OF~ 5 gci4icckow 3.I 3 APR 3 0 1993 4.3.A.2 ac vit Mar i o erable Co trol Rods d.DELETED f e.Sg S.l.~.(I'iieec,i'rc,4. 4$4 so~a 48, a.a~(o Co>>J 5'en D Co 44OE Control rods vith inopera le accumulator or t ose vhose position cannot be positively determined shall be considered noperable. hi 5 cvcd y 29 ho~a Inoperable control rods shall be positioned such that Specification 3.3.A.l is met In a tion ur ng, rea over o eration more than one control rod in any 5x5 array may be inoperable (at least 4 OPERABLE control rods must separate any 2 ino erable ones f this spec fication cannot be met e~reactor a sMrHd or at pover the reactor shall e brou ht to SHUTDO COHDITIO t n 4+hours.IZ.d.The control rod accumulators shall be determined OPERABLE at least once per 7 days by verifying that the pressure and level detectors are not in the larmed condition. sec.@~ED CtL+ko A~t rdqrS~~ptC g'C g'I 5 Prot os'o)<s.e.-aA;g.<0 g.o 0<d~2-4 V~k o4 1'.sc 0 Co 4.4'o P~psv0$bpM rcpt Ache~A),(y)'2 BFN Unit 2 3.3/4.3-4 AMENDMgP gp.2>2 r PsyQ>>I,

APR 3 0'l993 ds l.Each control rod shall be A4, coupled to its drive or corn lctel nserted and the control rod directional "<~~'~<~control valves disarmed A J ccr a1%This equ rement does n t apply n the S DOWN CO ITIOUS w en the r ctor is nted.'la control od arrives a be oved as ng as Speci cation 3.3.h.l is met A7 4 eaaf'S Ql+Ll~Q 4e~rg The coupling integrity shall be verified for each withdrawn control rod as follows: a.Verify that the control rod is folio ing th drive by obs ving I respons in the nuclear tru-entation ach ti e a rod is mo ed wh n the rea tor is crating ove the eset power level utoff of the RWM.sg.+I.3,5 b.When the rod is.fully withdrawn irst time after each re i outage o ter mainten observe that the drive does not go to thc ovcrtravel position.2.The control rod drive housing support system shall be in place during REACTOR POWER OPERhTIOH or when the reactor coolant system is pressurized above atmospheric with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3 3.h.l is met.2.The control rod drive housing support system shall be inspected after reassembly and the results of the inspection recorded.BFS Unit 2 3.3/4.3-5 AMENOMENT HO.2 y 2 I"AGE 0' o es 4.3.C.S e o s Sg 3.1.$."t X inserted From Avg.Scram Inser-5 20 50 90 0~398 0.954 2.120 3.80 3.The maximka scram insert on time for 90K insertion of any OPERABLE control rod shall not exceed 7.00 seconds.2.The average of the scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a two-by-two array shall be no greater than+2.At 16-week intervals, 10K of the OPERABLE contxol rod drives shall be scram-timed above 800 psig.Whenever such scram time measurements are made, an evaluation shall be made to provide reasonable assurance that proper control rod drive performance is being maintained. W<~7I FlcriTio hf Fad CIA PC W F44 BFN ISTS P.l,af D~D.v a e The reactivity equivalent of the difference between the actual critical rod configuration and the expected configuration during power operation shall not exceed 1%, hk.lf this limit is exceeded, the reactor will be placed in SHUTDOWH COHDITIOH until the cause has been determined and corrective actions have been taken as appropriate. During the STARTUP test program and STARTUP following refueling outages, the critical'od configurations will be compared to the expected configurations at selected operating conditions. These comparisons will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data.This comparison will be made at least every full power month.See a-uS tjFIC<VI49 RR"~"+<~roy urn IsrS Z.j.2 BFH Unit 2 3.3/4.3-11 >MENMENT No.y7'5 PAGE~pF

UNIT 3 CURRENT TECHNICAL SPECIFICATION IVIARKUP SP~ci+i~+g~ 'p APIt 3 0 1993 c-Co P.i.3 6eck CC C'Q((be eP6fh ELE ppCRABrli<t'aner~l.u Coatrol rod drives vhich can-not be moved'vith control rod drive pressure shall be considered inoperable. f a part ally or fully vith-drava control rod drive can-not be aoved vith drive or scram pressure the reacto!shall be brought to the 8.~MH CORDITIOH thin hours and shall aot bc Acmic n s ar ed unless I)d onstra d that cause of e fail e is ao a fai c trol r dri ae m ollet hous a equate RcQjg$t ovn%Or gin has been~4'on deaonstrated as required 4.3,A.2.~i,n 72 hA4f5 4.3.k.2 SRB.i.3.Z a.3 lo3.Ceecir~ikh'on P.3 Bach partially or fully vithdravn OFBRABLZ control rod shall be exerc se cmc notch at least once each e vhea operating above the pover level cutoff of the M.In the event pover operatioa is ccmtinu vith ee or aore iaoperablc contro rods this test 1 be perfonaed at least once ea day, vhen operating above the pover level cutoff of the NN.The control rod direc-~<)""~tioaal control valves for 1aonoroblo oontrol rods shall bc disaraed~>A'o'.z.b.DELl'.TED ru)oXg kaw.rcpt(~Hen Ctquirc4 Ash'un g.l/4l q BHt Unit 3 3 3/4.3-2 AMENDMENT ND.Z 6 9'AGE X GP 0! Cont'<6~nc~on C.(g c ntro rods ith cram ti s g ster han hose pe tte by Sp cifi ation 3.3.3 ar IHOP ut f ey can be inserted vith control rod drive pressure they need not be disarmed electrically. Co Shen t is initially determi d that a control o is incapable of rmal insertion a tes shall be conducte to onstrate that th ca e of the malf ction is not a failur in the control d drive mechanism. If this can be demo rated an attempt to lly insert the contr od shall be made.I the control rod c t be inserted and investigation has d trated that the caus of failure is not a f led control rod ive mechanism coll housing, a shutd rgin test shall e ma e to demonstrate und this condition that e core can be made su ritical for any reict ity condition d ing the emainder of e o rating cycle ith the lytically dete ned highest vorth c trol rod capable o vithdraval fully vithd, and all other cont 1 rods capable of insertion fully inserted.BFH Unit 3 3.3/4.3-3.FAGER~=(u

APR 3 0 1993~~~OPaqhu;~y 4'.3.A.2 ont'd C~N 3.l.g.~Control rods vith ino crabl ccumulator or those vhose position cannot be positively detezmiaed shall be coasidered ino crab e.5~$X'thony~d The control rod accumulators shall be determined OPERABLE at least once per 7 days by verifying that the pressure and level detectors are not in the alarmed condition. Rogu ted/@liens P.g S~L++1 MnCk4on f.inoperable control rods shall be positioned such that Specification 3.3.h.l is met.a t oa ng react povcr o eratioa no mo an onc control rod in any 5x5 array may be inoperable (at least 4 OPERABLE control rods must separate any inopcrab es)If this specification cannot S~5~HikC<Hoy 4>>~g<~kr frofnscd SPecig~mhcn Z.<.5 P<oP>sad]A~+>eh'hon 0 g.CO P 0 cond'Van)E be.cr c r bc st r d or if at pover thc reactor shall c brought HUTDO~hours.IZ.f fr.~~it)on frppused kpw,red AKP$5 P.g$g 2 BFH Unit 3 3.3/4.3W AMENPMENTgp ~69 3~ii APR 3 0 1993 l.Each control rod shall be coupled to its drive or c e 1 erced the pcÃpn control rod directio control valves d cd e t a.s oes t ap 1 the WS CO ITI en th rea or ia vent cont ol r driv Ra be rcaov as as Spe ifica on 3.A.l is R c.witkn hcu)5'~HL n 1 h4ufS Veri that e con ol ro is to oot Oho ioo obse ing capo c in c nucl r i tru-R aCi ea CiRc, a od Rov d cn c r ctor is erat ab ve pres po r vcl off the SR Z-i.8.5'.The coupling integrity shall be verified for each vithdram control rod as follovs: Sg g,(,3.5 b.Mcn the rod ia fully vithdravn thc irat t e Cer each refueling outage 0 ccr Raintenance, observe t the drive does not go to the overtravel sition.2o The control rod drive housing support syatca shall be in place during REACTOR.HNER OPERATION or shen the reactor coolant systca is pressurised above ateosphcric pressure vith fuel in the reactor vessel, unlesa all control rods arc fully inserted and Specification 3&.Aol is Let.2.The control rod drive housing support systea ahall be inspected after reasacably and the results of the inspection recorded.BFS Unit 3 3.3/4.3-5 AMENDMENT gp, y 6 g PAGE~OF

Qpi>pc'i4icaVion 3.1-3 NOV>8 taM 3.3.C.4.3.C.Sc I Inserted From Avg.Scram Inser-'t sc 5 20 50 90 0.398 0.954 2.120 800 3.Thc maximum scram insertion time for 90K insertion of any OPERABLE control rod shall not exceed 7.00 seconds.2.The average of thc scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a tvo-by-tvo array shall be no greater than: 2.At 16-vcek intervals, 10K of the OPERABLE control rod drives shall be scram-timed above 800 psig.Whenever such scram time measurements arc made, an evaluation shall be made to provide reasonable assurance that proper control rod drive performance is being maintained. ~s~g4cgWn Fi<charge>Po~BPrv xsvs 3-i-9 D.The reactivity equivalent of the difference betveen the actual critical rod.configuration and the expected conf iguration during povcr operation shall not exceed 1X 4k.If this limit is exceeded, the , reactor vill be placed in the SHUTDOWNÃCOHDITIOS until the caudle has been determined and corrective actions have been taken as appropriate. During the startup test program and startuy folloving re fueling outagcs, the critical rod configurations vill be compaicd to the expected configura'tions at selected operating conditions. These comparisons vill be used as base data for reactivity monitoring during subsequent pover operation throughout the fuel cycle.ht specific pover oyerating, conditicms, the critical rod configuraticm vill be compared to thc configuration expected based upon approyriatcly corrected past data.This comparison vill bc made at least every full pover month.See Tgyf)~hon ka<MgeS 4R Spy BPK Uait 3 3.3/4.3-11 AMENDMENT N.Z P 9 PAGE~OF~ l'(4 l JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3<<CONTROL ROD OPERABILITY ADMINISTRATIVE CHANGES Al All reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS)should be more readily readable, and therefore understandable, by plant operators as well as other users.During the reformatting and renumbering of the improved Technical'pecifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. The organization of the Control Rod OPERABILITY specification is proposed to include all conditions that can affect the ability of the control rods to provide the necessary reactivity insertion and also to be simplified as follows: 1)a control rod is considered"inoperable" when it is degraded to the point that it cannot provide its scram function, when decoupled, or when its position is unknown.All inoperable control rods (except stuck rods)are required to be fully inserted and disarmed.2)a control rod is considered"inoperable" and"stuck" if it is incapable of being inserted and requirements are retained to preserve shutdown margin for this situation. 3)a control rod is considered"slow" when it is capable of providing the scram function but may not be able to meet the assumed time limits.4)and special considerations are provided for conformance to the banked position withdrawal sequence (BPWS)at less than 10%of rated thermal power.The scram reactivity used in the safety analysis allows for a specified number of inoperable and slow scramming rods, and the control rod drop accident analysis provides additional considerations of the BPWS at low power levels.Two"Notes" have been added.The first Note (at the start of the ACTIONS Table)provides more explicit instructions for proper application of the ACTIONS for Technical Specification compliance. The Note allows separate Condition entry for each control rod.In BFN-UNITS 1, 2, 5 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY ADMINISTRATIVE CHANGES CONTINUED conjunction with proposed Specification 1.3,"Completion Times," this Note provides direction consistent with the intent of existing Actions for inoperable control rods.The intent is to allow a specified period of time, for each inoperable control rod, to verify compliance with certain limits and, when necessary, fully insert and disarm.The second Note, which is consistent with the requirements of proposed LCO 3.0.2, has been added to the ACTIONS and allows the RWH to be bypassed, if needed for continued operations, provided appropriate ACTIONS of proposed LCO 3.3.2.1 (RWH Specification) are taken.This is a human factors consideration to assure clarity of the requirement and allowance. A2 The requirement that control rods with scram times greater than those permitted by Specification 3.3.C.3 be considered inoperable (CTS 3.3.A.2.c) is included in proposed SR 3.1.3.4.The actions for control rods with scram times greater than the limit are more restrictive (see comment M4).Eliminating the separate Specification for excessive scram time by moving the requirement to another Specification, does not eliminate any requirements, or impose a new or different treatment of the requirements (other than those proposed in Comment M4).Therefore, this proposed change is considered administrative. A3 These requirements have been deleted since they are redundant to those currently found in BFN TS 3.3.A.2.a. Changes to that Specification are justified in the comments relating to that Specification. As such, this change is consider administrative. A4.This provision has been included in proposed BFN ISTS LCO 3.0;4" ("motherhood") and need not be repeated in individual Specifications. Proposed LCO 3.0.4 does not permit entry into a MODE or other specified condition in the Applicability except when the associated ACTIONS to be entered permit operation in the MODE or other specified condition in the Applicability for an unlimited period of time.Therefore, removing this requirement is considered an administrative change.BFN-UNITS 1, 2, 5 3 Revision 0 PAGE> JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY ADMINISTRATIVE CHANGES CONTINUED A5 The"shutdown condition" has been more accurately described as"hot shutdown condition", i.e., MODE 3 in the proposed BFN ISTS.This is a human factors consideration to clarify the intent since currently"shutdown" could mean either hot or cold shutdown based on the definition provided in BFN TS 1.0.A6 The requirement that control rods be coupled to their drive mechanism is covered by proposed SR 3.1.3.5;thus, making it a requirement for control rods to be considered OPERABLE.Eliminating the current separate LCO for control rod coupling, by moving the surveillance and actions to proposed BFN ISTS 3.1.3, does not eliminate any requirements, or impose a new or different treatment of the requirements (other than those separately proposed). Therefore, this proposed change is considered administrative. A7 This requirement duplicates an identical and more appropriately placed requirement in existing Specification 3.10.A.6.Therefore, deletion of this requirement is considered administrative. This Surveillance has been changed to more explicitly describe the requirement, which is to ensure that coupling is verified if maintenance on the control rod affected coupling.If maintenance is performed that does not affect coupling (e.g., HCU valve maintenance) there is no reason to perform testing.TECHNICAL CHANGES-MORE RESTRICTIVE Ml Proposed Required Actions A.2 and B.1 are comparable to CTS 3.3.'K.2.b, which requires inoperable control rods (including stuck control rods)to be disarmed.Two hours is allowed to disarm withdrawn control rods that are stuck.Since CTSs do not provide a maximum time limit, the proposed change is considered more restrictive. Proposed SR 3.1.3.2 and SR 3.1.3.3 require control rods to be inserted rather than the existing requirement of exercised, which could be met by control rod withdrawal. It is conceivable that a mechanism causing binding of the control rod that prevents insertion could exist such that a withdrawal test would not detect the problem.Since the purpose of the test is to assure scram insertion capability, restricting the test BFN-UNITS 1, 2,&3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY TECHNICAL CHANGES-MORE RESTRICTIVE CONTINUED to only allow control rod insertion provides an increased likelihood of this test detecting a problem that impacts this capability. This Surveillance has been moved to Required Action A.3.In addition, this is now required when as few as one control rod is immovable. Added Required Action C.l, which requires an inoperable rod (unless stuck)to be fully inserted within 3 hours and disarmed within 4 hours.Placed a time limit on existing TS 3.3.A.2.b for disarming control rods (Required Action C.2)and existing TS 3.3.B.1 for inserting and disarming control rods.This is more restrictive than current requirements, which allow the rod to remain withdrawn when inoperable. Also, this is more restrictive since the ISTS requires disarming even if rod can be inserted with drive pressure.Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.The control rod is disarmed to prevent inadvertent withdrawal during subsequent operation. Reference related Comment Al.Since existing Technical Specifications do not provide a maximum time limit, the proposed change is considered more restrictive. This requirement has been modified to require the position of each control rod to be verified every 24 hours (proposed SR 3.1.3.1).Current requirements do not have a specific Surveillance for this requirement. M6 Proposed Required Actions D.1 and D.2 allow 4 hours to restore compliance with the Specification (i.e., restore control rods to operable status or restore compliance with the BPWS).This change is considered more restrictive since the current time to reach a shutdown condition (MODE 3)has been reduced from 24 hours to 12 hours (per proposed Required Action E.1).Since the total time to reach a shutdown condition has been effectively changed from 24 hours to 16 hours (4 to restore and 12 to reach MODE 3), this proposed change is considered more restrictive. A new Condition has been added (second part of proposed Condition E)requiring an shutdown (i.e., be in MODE 3 within 12 hours)if 9 or more control rods are inoperable. Currently, 9 control rods can be inoperable, provided they are separated by four operable control rods, without requiring shutdown.BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY TECHNICAL CHANGES-MORE RESTRICTIVE CONTINUED H8 Proposed Required Action A.1 has been added to confirm that when a control rod is found stuck, it is properly separated from"slow" control rods.The other Required Actions of ACTION A were renumbered to reflect the insertion of A.l.The scram reactivity analysis assumes, among other things that there are two"slow" control rods adjacent to one another, a third control rod is stuck in the withdrawn position, and a fourth control rod fails to scram during the transient/accident analysis (the single failure).However, the analysis does not assume that the original stuck control rod is adjacent to the two"slow" rods or to another"slow" control rod.If this occurs, the local scram reactivity rate assumed in the analysis might not be met.H9~N10 Changed Frequency for verifying coupling to each time the rod is withdrawn to the full out position, not just the first time after each refueling outage.Existing Specification 3.3.A.2.f requires that inoperable (and stuck)control rods be positioned such that SDH requirements (3.3.A.1)are maintained. Proposed Required Actions A.4, B.2 and C.1 for LCO 3.1.3 requires that with one stuck rod (A.4)that shutdown margin be verified within 72 hours (Justification Ll), with more than one stuck rod (B.2)that the reactor be in Hot Shutdown within 12 hours, and with one or more inoperable rods (C.1)that each inoperable rod be fully inserted.By allowing one stuck rod and by requiring that all insertable control rods be fully inserted, the proposed Required Actions provide greater assurance that SDH is maintained than the requirement for verifying SDH for multiple rods withdrawn. Hll The current time to reach a non-applicable condition has been reduced from 24 hours to reach Cold Shutdown (MODE 4)to 12 hours to reach MODE 3 (per Required Action E.1).This change is more restrictive because all rods must be fully inserted in 12 hours instead of the currently required 24 hours.Cooling the unit down (proceeding from MODE 3 to NODE 4)does not provide any additional margin and, in some cases, could be counter productive since positive reactivity is inserted during cooldown.BFN-UNITS 1, 2, EL 3 Revision 0 PAGE~OF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Details of the methods of disarming control rod drives (CRDs)are relocated to the Bases and procedures. The.requirement to disarm the CRD remains in the Specification. LA2 Details of the methods of verifying control rod coupling are relocated to plant procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll Proposed Action A allows continued operation with one withdrawn control rod stuck provided that Shutdown Margin is demonstrated. With a single control rod stuck in a withdrawn position, the remaining control rods are capable of providing the required scram and shutdown reactivity. Failure to reach COLD SHUTDOWN is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain HOT SHUTDOWN conditions. Required Action A.3 of LCO 3.1.3 performs a notch test on each remaining control rod to ensure that no additional control rods are stuck.The reason for the failure (e.g., failed collet housing)is not significant provided all other rods are tested to ensure a like failure has not occurred.Given these considerations, the 72 hours allowed to demonstrate SHUTDOWN MARGIN is considered reasonable to perform the analysis or test.L2 Proposed SR 3.1.3.3 extends the surveillance that verifies control rods are not stuck from 7 days to 31 days for control rods that are not fully withdrawn. This is consistent with the BWR Standard Technical Specifications, NUREG-1433. Partially withdrawn control rods have a significantly greater effect on core flux distribution than do fully withdrawn control rods.Historically, power reductions are required each week to perform the test on partially withdrawn control rods.The impact of testing on plant capacity is deemed excessive given the following considerations: PAGE BFN-UNITS 1, 2, 5 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY TECHNICAl CHANGES-LESS RESTRICTIVE CONTINUED 2)At full power a large percentage of control rods (typically 80-90/)are fully withdrawn and would continue to be exercised each week.This represents a significant sample size when looking for an unexpected random event.Operating experience has shown that"stuck" control rods are an extremely rare event while operating. 3)Should a stuck rod be discovered, 100%of the remaining control rods (even partially withdrawn) must be tested within 24 hours (proposed Required Action A.3).L3 The requirement that no more than one control rod in any 5 x 5 array may be inoperable (at least four operable control rods must'eparate any two inoperable ones)is proposed to be changed to allow inoperable control rods to be separated by two operable control rods.This is consistent with the safety analyses associated with this limitation. Proposed ACTION D addresses the condition when the reactor is x 10%RTP and two or more inoperable control rods are not in compliance with the BPWS and not separated by two or more operable control rods.The required action is to restore compliance with the BPWS within 4 hours or restore the control rod to operable status within 4 hours.Inoperable control rod separation requirements are required at a 10%RTP because of Control Rod Drop Accident (CRDA)concerns related to control rod worth.Above 10%RTP, control rod worths that are of concern for the CRDA are not possible.The proposed two operable control rod separation criteria in ACTION D is acceptable for the BPWS analysis and therefore, is acceptable for use in the proposed TS.L4 The current TSs require a daily notch test in the event power operation is continuing with three or more inoperable control rods and the plant is operating at>30%RTP.The proposed TS only require the control rod notch test in the case of a single stuck control rod, and only once within 24 hours.The purpose of the control rod notch test on each withdrawn operable control rod is to ensure that a generic problem does not exist and that control rod insertion capability remains.The single performance of the control rod notch test satisfies the same function as the daily notch test of the current TS without requiring the additional testing.BFN-UNITS 1, 2, 8, 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.3-CONTROL ROD OPERABILITY TECHNICAL CHANGES-LESS RESTRICTIVE CONTINUED L5 The requirement (control rod separation requirement) associated with the proposed Note to Condition D (which limits the requirement to a 10%RTP)is necessary to ensure the rod pattern is in compliance with BPWS.This ensures that a rod drop accident will not result in excessive local power in a fuel bundle.Analysis has shown that inoperable control rod distribution is not a problem when>10%RTP.The analysis is described in NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel," Revision 8, Amendment 17.This analysis also showed that the inoperable control rod distribution is needed at x 1%RTP, which is broader than the current requirement for reactor power operation. The inoperable control rod distribution requirement has been modified to include this new restriction. Therefore, any decrease in safety by eliminating the distribution requirement >10%RTP, is offset by the added safety of requiring inoperable control rod distribution at lower power when a rod drop accident can impact fuel design limits.t RELOCATED SPECIFICATIONS Rl CRD OPERABILITY re quirements (CTS 3.3.B.2)currently include requirements for the CRD housing support to be in place.These requirements have been relocated to plant procedures. The CRD Housing Support does support CRD operability which is part of the primary success path.Having the CRD Housing Support out of place does impact CRD operability. It is indirectly covered in ISTS 3.2.3 in the blanket action for a control rod being inoperable for any other reason.There is no need to duplicate requirements in a subsystem LCO.Relocation of this LCO is appropriate since plant configuration (the control rod housing support in place)wold be control by post maintenance procedures. BFN-UNITS 1, 2, 5 3 Revision 0 f'ACiE~OF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP ~~.rPtk A~'L tT%0tVQVr~~~Vii'L avIa'rvrvIit aa4~~I rr rL eteail 1~1 XJ 4%~+a~all e, cvr~.~.r s tt v ef~~~\'~~~'~~QS~~~~~~~~~~~~~~~~~~~~~~.'~~~~~~~~~~~~~4~~~~~~I~~~'~~~~~~~~~I~~I~~~~~~~I~~I I~~~~~~~I'~~I~~ Qci PiQzhon p J,g NOV 0 2.The avcragc of thc scram inser-tion times for thc three fastest operable control rods of all groups of four control rods in a tvo-by-tvo array shall be no greater than: Z Inserted From hvg.Scram Inser-t S 5 20 50 0.398 0.954 2 120 3.800!3.The maximum scr inserti,on time for 90'nse ion of opera c control ro shall not exceed.00 seconds.SR s.i.q.i 2<>At-week 10 of thc operable contro rod drives shall bc scram-c above 800 psig.cnevcr su scram m dsurem ts are made g an ev luatio shall e made to ovide rcaso le assu ce t t pro cr contr rod rive perfo ce i bef intained sR.>.9Z AlO S lF r, l.q, D.D.Thc reactivity equivalent of the diffcrencc betvecn the actual critical rod configuration and the expected configuration during power operation shall not exceed 1X hk.If this limit is exceeded, thc reactor vill be placed in the SHUTMWK COMITION until the cause has been determined and corrective actions have been taken as appropriate. ~4.ch Q>>PrqoSed gP'Q)STD p)~During the startup test program and startup following refueling outagea, the critical rod configurations vill be compared to the expected configurations at selcctcd operating conditions These comparisons vill be used as base data for reactivity monitoring during subsequent pover operation throughout the fuel cycle.kt specific pover operating conditions, the critical rod configuration vill be compared to the configuration expected based upon appropriately corrected past data.This comparison vill be made at least every full pover month.BPK Unit 1 3.3/4.3-11 >~'.ENOMENT NO.I 7 2 0~ GEC 07 1994 4.3.E/}<nod If Specifications 3.3.C and.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the WSHUTDOWN CONDITION within Cours.Su eil anc requi erne s a eg a spe ifi d in 4 3.C and D a ove 3-3.F 1.The scram discharge volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.l.a.The scram discharge volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter. The valves may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation. l.b.The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MM.2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.3.If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBX CONDITION within 24 hours.~2~3.When it is determined that any SDV drain or vent valve is.inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE immediately and weekly thereafter. No additional surveillance required.+~S~iVlaok'ow +~C~~ps,>~4 I$T5 3.I~f'AGE~"'~BFN Unit 1 3.3/4.3-12 AMENOMENT N.g g~ UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

APR 8 0 1993 gg(<ni~ale.3.1'/-l 1'E.iA-t)a'CA.Sc 1./Co 3.l,g am se tion Times t on imes sec 5 20 50 90 0.375 0.90 2.0 3.500 The average scram insertion time, based on the deenergization of the scram pilot valve sole-noids as time zero, of all OPERABLE control rods in the reactor pover o eration conditio hall be no greater than: Pro~>/N4 d'or SRg O 1.A refueling a OPERABLE rods shall be scram-time tested from the fully vithdravn position vith the 85 nuclear system'pressure 800~psig.This testing shall be completed rior to exceeding 40K!po elov pover, onl rod in t se sequ ces ich we fully thdra in t e region from 100K d densi to 50K, rod ensi,ty shall be scram-ime tested.2dl Fiiye ay g p~p's ca SR 3./.0, I~r46z.J,Q-I ~gq~/BFH Unit 2 3.3/4.3-9 AMENDMENT HO.2 I 2 PAGE Al 5 Inserted From Avg.Scram Inscr-5 20 50 90'.398 0.954 2.120 3.800 2.The average of the scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a tvo-by-tvo array shall be no greater than: o e sa~.l.4~2.At 6-week intervals 10K of thc OPERABLE control g.l rod drives shall be scram-timcd abovs 800 psig.encvcr su scram time me surem ts are ade, an cva uation shall mad to p vide easonab e assur ce th prope control rod dr ve performance is being maintained SR 3.l.g, sR.me for 90K crtion any OP LE contro rod shal&not excee 7.00 seconds.D v o a The reactivity equivalent of the difference betveen the actual critical rod configuration and the cxpccted configuration during pover operation shall not exceed 1X hk." If this limit is exceeded, the reactor vill bc placed in SHUTDOWH COHDITIOH until the cause has been determined and corrective actions have been taken as appropriate. F'~c47io CHAhlC8$p'og pgopq~~~PPN lares 3 I Z During the STARTUP test program and STARTUP following refueling outages, the critical rod configurations vill bc compared to the expected configurations at selected operating conditions. These comparisons vill be used as base data for reactivity monitoring during subsequent pover operation throughout thc fuel cycle.At specific pover operating conditions, thc critical rod configuration vill be compared to the configuration expected based upon appropriately corrected past data.This comparison vill be made at least every full pover month.BFH Unit 2 3.3/4.3-11 NENMENT NO.yeas CII;'.",-3 o&H9%%TO>%r 0 L'OO meoo'11 rttie aern lir v t aisa nsarjiar)or(eov taio~~~~~f~~I~~~~II~~~~~~~I:~~I I~~I~I(I~~~I((I~I~~I I I~I~~.O'.I~II~~~~~~~~~~'~'I~~II~~'~~~,~~~~~~~~~~I~~II~~~I~~'~~I~'~~~~~~~I~~~~I I~~~~II~~~~~~~~'I A~~I~.~~~~~~~~~~~~~so~~~~~~~~~~,~~~~~~a~I I I'8'I I~~~~~~~~~I~'~~~~~~~o~~~~~~;~Ils I~~~~

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP

~~~I~'~~T%'W'J.'I A'hV'a J~J'i~'~DI~~~~~':~~~~~~~~~~~~~~~~~~~~~~~~~~~~~4'h~~~~~~~~~~.I~~~~~~~~~~~~~~J I~~~~~~~I~~~~~~~~~4 J'I I g 1 o NOV I8 ISN average of the scram inser-tion times for the three fastest OPERABLE control rods of all groups of four control rods in a tvo by-tvo array shall be no greater than: 5 20 50 90 0.398 0.954 2.120 3.800 The ianm ram i crti time r 90%1 ertio of contro rod s ll n eed 7~sec I Inserted From Avg.Scram Inser-Qgg te/~2.2e h-vc 10K of thc OPERABLE control rod drfvcs shall be scram-L~timed Amvc 00 pslg.vcr scram casu cat arc dc an c lust on s ll b ma e to rovi e rc onab e ass ance that rope cont 1 r drlv perfo ce s be ng int cd.SR 3.i.9.~SR 8.l.q.Q D.D The reactfvity equivalent of the differeace betveen the actual critical rod configuration and the expected configuration during pover operation shall aot exceed 1X hk.It thfs lisLft is exceeded, the reactor vill be placed ia the SHDTDORf COSDITIOK uatil the ca6ie has beea deteraiaed aad corrective actions have beea takea as apyropriate. During the startup test program aad startup folloving refueling outages, the crf tical rod configurations vill be coapaied to the expected configurations at selected operating conditions. 'hese coaparfsons vill be used as base data for react fvf ty aoaf toring during subsequent yover operation throughout the fuel cycle.At syecific yover operating conditions, the critical rod configuration vill be compared to the conf iguratioa expected based upon apyropriately corrected past data This comparison vill b>>made at least every full pover month.5'e4-'C5QPi'egg'o~ gp CQ+<t'<OPoSed Df=g y5 Ty 3 BPÃQait 3 3.3/4.3-11 NENtmee.Z p g PAGE~OF I

~~QC7lo~(3.3.F.If Specifications 3.3.C and 3-3.D above cannot be met, an orderly shutdown shall be initiated and th reactor~shall be in th SHUTDOWN~~CONDITION within~hours.4.3.E.0 7 1994 S rve'an e requirements re a sp cifij'd in 4.3.C nd 4.3.D above.1.The scram discharge volume drain and vent valves sha11 be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.l.a.The scram discharge volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter. The valves may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation. l.b.The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MM.2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.2~When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE immediately and weekly thereafter. 3.If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBY CONDITION within 24 hours.3.No additional surveillance required.BFN Unit 3 See 3NHg'~g>~~gh 6r 8~~GE~GF~AMDtPMBP go.y 8 6

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.4-CONTROL ROD SCRAM TIMES ADMINI STRATI VE CHANGES Al Reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS)should be more readily readable, and therefore understandable, by plant operators as well as other users.During the reformatting and renumbering of the improved Technical Specifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. A2 CTS lists the position of the control rod in terms of%inserted from the fully withdrawn position.Proposed BFN ISTS Table 3.1.4-1 list the position in terms of notch position.These positions are equivalent (to the next nearest measured notch position)except expressed in different terms.A3~A4 The Surveillance Frequency has been modified to require testing after fuel movement within the reactor pressure vessel.This is equivalent to after each refueling outage, which implies that fuel has been moved.Proposed Specification

3.1.4 retains

the current maximum scram time requirement (7 seconds)of Specification 3.3.C.3 for the purpose of defining the threshold between a"slow" control rod and an inoperable control rod.Note 2 of proposed Table 3.1.4-1 ensures that a control rod is not inadvertently considered"slow" when scram time exceeds 7 seconds.Proposed SR 3.1.3.4 verifies each control rod's maximum scram time is c 7 seconds or it is considered inoperable. A5 CTS 4.3.C.1&2 requires scram time testing to be performed at>800 psig.SRs 3.1.4.1&2 require testing to be performed at a 800 psig.The requirement to perform this testing at pressure=800 psig is slightly less restrictive since the SRs can be performed over a slightly broader pressure range.However, since the change is so minor it has been categorized as administrative. The proposed change is consistent with BWR/4 Standard Technical Specifications (NUREG-1433). TECHNICAL CHANGES-MORE RESTRICTIVE The LCO for Control Rod Scram times ensures that the negative scram reactivity assumed in the DBA and transient analysis is met.Current BFN Unit 2 Technical Specifications accomplish this by specifying the I BFN-UNITS 1, 2,&3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.4-CONTROL ROD SCRAM TINES TECHNICAL CHANGES-NORE RESTRICTIVE CONTINUED maximum individual scram times (7.0 seconds), average scram times and local scram times (four control rod group).The design basis transient analysis assumes all control rods scram at the same speed.If all control rods scram at least as fast as the analytical limit, the scram reactivity assumed in the DBA and transient analysis is met.A distribution of scram times (some slower and some faster than the analytical limit)can also provide adequate scram reactivity. The more control rods that scram slower than the analytical limit, the faster the remaining control rods must scram to compensate for the reduced reactivity of the slower control rods.Proposed BFN ISTS 3.1.4 incorporates this principle to ensure adequate scram reactivity by specifying scram time limits for individual control rods instead of limits on average or four control rod groups.This methodology is similar to that being used for the BWR/6 STS.The LCO scram time limits have margin to the analytical scram time limits to allow for a specified number and distribution of slow control rods, a single stuck control rod and an assumed single failure.The proposed LCO specifies the number and distribution of"slow" control rods allowed that will still ensure the analytical scram reactivity assumptions are satisfied. If the number of"slow" rods is excessive (>13)or do not meet the distribution requirements, the unit must be shutdown.This change is more restrictive since the proposed individual times are more restrictive than the average times.Currently, the"average" time of all rods or a group can be improved by a few fast scramming rods, even when there may be more than 13"slow" rods.The proposed specification limits the number of slow rods to 13 and ensures each slow rod is separated by two operable rods.Table 3.1.4-1 is modified by a note (Note A), which state that control rods with scram times not within limits of the table are considered slow and those with times greater than 7 seconds are considered inoperable as required by SR 3.1.3.4.In addition, a note has been added to the Surveillance Requirements Table requiring that, during a single control rod scram time Surveillance, the CRD pumps be isolated from the associated accumulator. This ensures that accumulator pressure alone is scramming the rod, not the CRD pump pressure (which can improve the scram times).BFN-UNITS 1, 2,&3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.4-CONTROL ROD SCRAM TINES TECHNICAL CHANGES-NORE RESTRICTIVE CONTINUED Proposed BFN ISTS 3.1.4 applicability of NODES 1 and 2 includes power levels x 1%RTP when first pulling rods to go critical.The applicability for current TS 3.3.C.1 of"in the reactor power operation condition" is defined by CTS Definition 1.0.H as any operation in the STARTUP/HOT STANDBY or RUN MODE with the reactor critical and above 1 percent rated power.Therefore, the proposed applicability is more restrictive. M3 Added a.Frequency for performing scram time tests on all control rods prior to exceeding 40%RTP.This Frequency requires these tests after each reactor shutdown a 120 days regardless of whether refueling occurred.M4~N5 Added Surveillance Requirement (SR 3.1.4.4)that requires a scram time test after work on a control rod or CRD that could affect the scram time.The Surveillance requires a scram time test after reactor pressure has reached e 800 psig and prior to exceeding 40%RTP.CTS require the unit to be placed in the SHUTDOWN CONDITION (reactor in shutdown or refuel mode)if the specified limit is exceeded.CTS 3.3.E requires an orderly shutdown to be initiated and the reactor be placed in the SHUTDOWN CONDITION within 24 hours.Proposed BFN ISTS is more restrictive since it requires the unit to be placed in NODE 3 (Hot Shutdown)within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach NODE 3 from full power conditions in an orderly manner and without challenging plant systems.Therefore, the proposed change is considered acceptable. Added Surveillance Requirement (SR 3.1.4.3)that requires a scram test after work on a control rod, or CRD that could affect the scram time prior to declaring the control rod OPERABLE with reactor steam dome pressure<800 psig.This scram test is performed with the rod inserted and the accumulator drained and isolated.The head of the water from the reactor acting on the under piston area will cause the rod to insert to overtravel and demonstrate that the scram valves open and the scram discharge volume exhaust path is open.This shows that the rod can be scrammed without subjecting it to unnecessary stress.BFN-UNITS 1, 2, 5, 3 Revision 0 PAGE 0 0 H r JUSTIFICATION FOR CHANGES BFN ISTS 3.1.4-CONTROL ROD SCRAM TINES TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl CTSs allow only rods in those sequences which were fully withdrawn in the region from 100%rod density to 50%rod density to be scram-time tested when below 10%power.This ensures that in-sequence fully withdrawn control rods are tested at low power where most rod worth is a concern.The Rod Patten Control Specification and RWH ensure proper CR sequences are followed.Details of the restrictions, methods and purpose of the Surveillance are relocated to plant procedures. The requirement to perform scram time testing remains in the surveillance. LA2~LA3 Proposed SR 3.1.4.2 requires a"representative sample" of control rods to be tested each 120 days of operation instead of the currently required 10%of the OPERABLE control rods (CTS SR 4.3.C.2).The proposed change adopts the position of the BWR Standard Technical Specifications, NUREG 1433, that these details be located in plant procedures and summarized in the Bases for the Surveillance. Details of the method to perform or the purpose of the Surveillance are relocated to plant procedures. The requirement to perform scram time testing remains in the surveillance. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll Proposed SR 3.1.4.2 is performed at 120 days cumulative operation in NODE 1 versus the CTS requirement of 16-week intervals. Sin'ce the proposed frequency is longer than 16-weeks it is considered less restrictive. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle.This Frequency is reasonable based on the additional Surveillances done on CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5,"Control Rod Scram Accumulators." PAGE Revision 0 UNIT 1 CURRENT TECHNICAL SP ECIFI CATION MARKUP 0 e APR 80 Sy Lco Z,1<5 d.DELElXD 5cC YQSN~I4~h Ap 4m~+g 8PN l 5'75 g,~,p propiSecp Acr'imam Ap8 CtP f.Inoperable control rods shall be positioned such that Specification .3.3.k.l is met.In addition, during reactor pover operation, no more than one control rod in axe Sx5 array may be.inoperable (at least 4 OPERhBLZ control rods must separate any 2 inoperable ones).If this specification canno be met the reactor shall not be started, or if at pover, the reactor shall be brought to a SHUTDOWN COHDITIOH vithin 24 hours.e.Control rods vith inoperable acc 1 ors ose v os osition cannot be itivel determine shall be considered inoperable. .1.5}e.: rn d~control rod accumulatorg~hhk4~ l 7 d PgS~ee<s P'Q BFH Unit 1 3.3/4.3W hMENDMBlT gP.y 9 6 Pfs.' UNIT 2 CURRENT TECHNICAL SP ECIFI CATION MARKUP PAGE~OF~

~QA<rue@~st:C'c.4'o Z.CO 9./5 4'or 04~yts<or DELETED GFN isrs 2~3 SR'3.I,S;I 5 eciV>'4aAoq 3/s APR 8 0 TOSS ods~~)~.g each d.~control rod accumulator/ eh~~I L.I pr op+St~A cvjanG A~3, Ce Q e.Control rods vith ino erable accumulator or those vhose position cannot be ositively determine shall be considere noperable. f.Inoperable control rods shall be positioned such that Specification 3.3.A.l is met.In addition, during reactor pover operation, no more than one control rod in any Sx5 array may be inoperable (at least 4 OPERABLE control rods must separate any 2 inoperable'nes). If this specification cannot be met the reactor shall not be started, or if at pover, the reactor shall be brought to a SHUTDOWN COHDITIOH vithin 24 hours.once per 7 day PrtSSiil C iS~'FVO PS iP BFN Unit 2 3.3/4.3-4 e@DMEjltt x0 z Z z.

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF 0 APR30m L(,o 3.).5 d.DPZZXED (Cont'd)A vcti4)~~gg SQ3-l.5..~control rod accumulator/ ha~be-I Il'Pr os)AB'i'd&Pp g~c+0 e.Ccmtrol rode vith inoperable a ator or oae vhoe position cannot be poaiti ely detcrmin shall be c ared inoperable. cmce per 7 days f.Inoperable control rods shall be positioned such that Specification 3.3.k.l ia met.In addition, durfng reactor pover operation, no aore than one control rod in any Sx5 array may be inoperable (at least 4 OFERDKZ control rode must separate any 2 inoperable ones).If this specification cannot be met the reactor shall not be started, or if at pover, the reactor shall be brought to a SHUTDOWNS COHDITIOI vithin 24 hours.PsS'~iS m qeo ps('5<e 3u5Hg;cab'on $i than~/en BOW)S~s r.),3 BFH Unit 3 3.3/4.3-4 AMENOMEQ go, 169 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.5-CONTROL ROD SCRAM ACCUMULATORS ADM IN I STRATI VE CHANGES Al Reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS)should be more readily readable, and therefore understandable, by plant operators as well as other users.During the reformatting and renumbering of the improved Technical Specifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. A2 Proposed SR 3.1.5.1 requires that the accumulator pressure be checked to ensure adequate accumulator pressure exists to provide sufficient scram force.This satisfies the intent of the existing surveillance. Therefore, the proposed changes are considered administrative. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic"~LAI Details of the method to perform or the purpose of the Surveillance are relocated to plant procedures. The requirement to ensure adequate scram pressure exists, to provide the necessary scram force, remains in the surveillance. The primary safety concern is accumulator pressure.Increasing water level indicates deterioration of the accumulator piston seal to the nitrogen side.The requirement for verification that the level detectors are not in alarm has been relocated to plant procedures. Changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll Proposed BFN ISTS 3.1.5, which replaces BFN TS 3.3.A.2.e, allows a short out of service time for the accumulators (Actions A and B also allow the control rods to be declared"slow" instead of inoperable) prior to declaring the associated control rods inoperable provided that proposed-ACTIONS A, 8, C and 0 are met.The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod scram accumulator. This is acceptable since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable accumulator. Complying with the Required Actions may allow for continued operation and subsequent inoperable accumulators governed by subsequent Condition entry and application of associated Required Actions.BFN-UNITS 1, 2,&3 Revision 0 Q!0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1e5-CONTROL ROD SCRAN ACCUMULATORS TECHNICAL CHANGES-LESS RESTRICTIVE CONTINUED Proposed Action A allows one control rod scram accumulator to be inoperable for up to eight hours when reactor steam dome pressure is~900 psig before declaring the associated control rod s'cram time slow or declaring the associated control rod inoperable. With one accumulator inoperable, the control rod may be declared"slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the required scram times.Since the existing action (BFN TS 3.3.A.2.c) to declare the control rod inoperable would allow the control rod to remain withdrawn and not disarmed, the proposed action to declare the control rod"slow" is essentially equivalent. The proposed limits and allowance for numbers and distribution of inoperable and"slow" control rods (found in proposed LCOs 3.1.3 and 3.1.4 respectively) are appropriately applied to control rods with inoperable accumulators whether declared inoperable or"slow." Required Action A.1 is modified by a Note indicating that declaring the control rod"slow" only applies if the associated control scram time was within the limits during the last test.Proposed Action 8 allows two or more control rod scram accumulators to be inoperable for one hour when reactor steam dome pressure is~900 psig provided charging pressure is restored within 20 minutes.Condition 8 requires that Required Action B.1 be taken in conjunction with Required Action 8.2.1 or 8.2.2.Required Action B.1 addresses the situation where additional accumulators may be rapidly becoming inoperable due to loss of charging pressure (charging pressure must be restored within 20 minutes).Required Actions 8.2.1 and 8.2.2 require that the associated control rods be declared"slow" or inoperable within one hour, which provides a reasonable time to attempt investigation and restoration of the inoperable accumulator. Since reactor pressure is adequate to assure the scram function and charging pressure is adequate, the proposed 1 hour extension is not significant. Proposed Action C allows one or more accumulators to be inoperable with*reactor steam dome pressure~900 psig provided that Required Action C.1 (verify that all control rods associated with inoperable accumulators are fully inserted)is taken immediately upon discovery of charging water header pressure (940 psig and Required Action C.2 (declare the associated control rod inoperable) is taken within one hour.Required Action C.1 must be completed immediately since adequate scram pressure is not guaranteed (i.e., reactor steam dome pressure~900 psig).BFN-UNITS 1, 2,&3 2 pAGE 2 OF~Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.5-CONTROL ROD SCRAN ACCUNULATORS TECHNICAL CHANGES-LESS RESTRICTIYE CONTINUED Once verification of adequate charging pressure is made (20 minutes is provided)and considering reactor pressure is adequate to assure the scram function of the control rods with inoperable accumulators, the proposed 1 hour completion time is not significant. In additions, since the reactor pressure may not be adequate to scram the rods in the proper time, Action C does not allow the rods to be declared"slow" (as allowed by Actions A and B).Proposed Action 0 requires an immediate scram if any Required Action or associated Completion time can not be met.This ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e., scram)of the control rods.This Required Action is modified by a Note stating that the action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed. BFN-UNITS 1, 2, 5 3 PAGE 3 OF~

INSERT PROPOSED NEW SPECIFICATION =3.1.6 Insert new Specification 3.1.6,"Rod Pattern Control," as shown in the BFN Unit 2 Improved Technical Specifications. JUSTIFICATION FOR CHANGES BFN ISTS 3.1.6-ROD PATTERN CONTROL TECHNICAL CHANGES-LESS RES RICTIVE Ll A specific requirement for control rods to be in compliance with the BPWS during oper ation at low power is proposed as TS 3.1.6.This proposed specification also contains an allowance (Actions to LCO 3.1.6)for,a limited number of out-of-sequence Operable control rods, which is presented in the BWR Standard Technical Specification, NUREG-1433, and also proposed to be included in the revised Technical Specifications. The Actions allow up to 8 out-of-sequence operable control rods (separate from any inoperable out-of-sequence control rods)to be returned to their correct position within 8 hours.This allowance for correction is proposed in recognition of the occurrence of such events as"double-notch" rod withdrawals, and minor misalignment of rod pattern during CRD hydraulic transients (control rod drift due to excessive cooling water pressure)or during a plant shutdown.These events can introduce out-of-sequence control rod patterns which the RWH was unable to preclude, even though the RWH was functioning as designed.BFN-UNITS 1, 2, 5 3 Revision 0 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 0 OV Zg lSBS F 4 kpyli to the oyexa status of th Standby Liquid ontrol System Applies to th sarvcillanc requirements o the Standby Liquid Control S tern.Qhur~v To assure e availability o a system Wth cayability to shnt dovn th xeactor and main in the shatdovn tion vithoat the ase of con ol rods.To veri<<y the op crab icy 0 f c St by tJ.quid Contro System.Rl LCO g,),7 44JA 4.Except as s'pecif ed 3.4.3.1, the Standby Liqa'4 Control System shall be OPERiSLZ at times vhcn re s el in the reactor vessel and the xeactor is not in a ahatdcwn condition vith Syecification 3.3.4.1 at edo v 41 14).Th oyer ility f the tandby Li id Co trol Sy tern s abc-veri ed the pe doorman of the foll tes 1.Verify yaay OPERABZLIT accordance vich Syecification 1.0.%f.2.it least once 4uring each oyerating cycle: a.e that, the cc='ng th sys cm t c ves is 1, 25.S WFl Unit 1 3.4/4.4 1 Manual y itiate:.".e stea, except cxplo-s ve va ves.V ual'y v ify v by p ping bo sol ion th oug'" the tecir" at'on't.". and adc to hc St dby Llqai Contr Solar'n aak.After pi..g b on lut'on, the sy cm hall be'ushc6 vit dern c vst ex'c t f y ninimum SR E,t,q.a AMENDS"sf'(T Mo.g~9 PAGE~OF S 0 vol<<<<SEP 02 1888 gp~c4c<yon S,).7 g 1<<o 5'R 3,(.q,(yumy flov rate of 39 gym against a svst m heact of'i ump~cr<<<<d va er tr che St by L quid C t=Tee Tank SR B.L 77 c initiate one of the Standby Liquxd Control System loo s and to the reactor vessel.s st ch eke los on of e ch zgc as ocia ed vi the tes d l oy, pr yer op tion f thc valv s o erabil ty e'y ac cs shall be selected such that the age of charge in service shall not exceed five years from che manufacturer's assembly date 5g'P['7'7 Il, Boch$TsccD5 t1lh~U4+g oth losivc valMs shal e tested in:he l A>course of tvo oycrating cycles.Unit 1 3 4/4.4-?hMENDMENT NO.Z g 4/V

<ec n3(7 FEB 2 S 1995 From and after the date that a redundant component is made or found to be inoperable, Specification 3.4.A.1 shall be considered fulfilled and continued operation permitted provided that the component is returned to an operable condition within seven days.1.Whe a c mponen is found to e ia perab, it re und t comp nent hal b dern strat d to e o era e imm diate y an il there fter til he inop rable ompon nt i repaired.LCo Rl l~~~5<3,>.g.q I Sg~,1.q,g At all tune ea e an y Liquid Control System is required'o be OPERABLE, the following conditions shall be met: 1.At least 186 pounds Boron-10 must be stored in the Standby Liquid Control Solutioa Tank and be availablc for injection. The sodium pcntaborate solution concentration must be equal to or less thaa 9.2L by weight.The oil ing ts ss e perf ed to ve ify the avail ili y of e Liq'd Contro S 1 ion.SR 3 I'7il~1.Volume: Check at least once pcr day.Sodium Pentaborate ],1 3 Concentration check at least once per month.Also check concentration within 24 hours anytime water or boron is added to the solution.SRS.i~V 3.Boron-10 Quantity:VC,<i Q+4+Co+Cea~jiow Q6d 4cw1perR4/1g W Qc oe i 1$olMQO+~LO,'WA L'w,a o<P;~p)g I/Hd Pi q<p(q (At least once per month, calculate and record the quantity of Boron-10 stored in thc Standby Liquid Control Solution Tank.SR3,I~~0 4.Boron>>10 Enrichment: At least once per 18 months and followiag each addition of boron to the Standby Liquid Control Solution Tank: fo p0$4Gt 5 R3.l.1,~BFN Unit 1 3.4/4.4-3 emw~N0>>>PAGE-OF i 5 ice DEC 07 1g94 a.Calculate thc cnrich-mcnt vithin'4 hours.b.Verify by analysis vithin 30 days.3.n 4'.D Sg 3.l.v 5 System conditions must satisfy the following equation.y 1 (13 wt.1)(86 gpm)(19.8 atom%)vhere, gg 3,]o 7~S Verify that the equation given in Specification 3.4.D is satisfied at least once per month and vithin 24 hours anytime water or boron is added to the solution.C~sodium pcntaborate solution concentration'veight percent)Dctcrmined by the most recent pe ormance of t e surveillance inst ction requi d by S ecif ation 4.4.C.2 Q~pump flov rate (gpm)termine y the most recent perf cc of the urvcillanc instru tion required y Specifi tion 4.4.A.2.E~Boron-10 enrichment (atom percent Boron-10)Det rmined by the most recent perf ce of the urveillance instru ion required y Specification 4.4.C.4.t l./@7~AD 84C If Specification 3.4.A through 3.4.D cannot be mct, make at least one subsystem OPERABLE within 8 hours or the reactor shall be placed in a SHUTDOWN CONDITION with all operable control rods fully inserted vithin the following 12 hours.1.No dditi surveillance re uired.BFN Unit 1 3.4/4.4-4 AMENDMENT NO.2 I 3 pAGE~ 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE Spec,'Aqua 3,/7 NOV 22 1988 3.4 4'Applies to the perating status of the Standby L quid Control ystcm.Applies o the surveian requircm ts of the Staadb Liquid Can"ol System.Eia~v To as re the availabili of a system th the capability to shut dova the reactor and ma tain the shutdo condition vithou the use of c trol rods.o verify the op"ability of the S dby Liquid Co trol System.lC0 RI7 Except as specified Qx 3.4.B.1, the Standby Liquid Control System shall be OPERABLE at ll t es vhen ere is fuel in the reactor vessel and the reactor is aot in a shutdovn condition vith Specificatioa 3.3.A.1 atisfied h.va~s abc The rability of e Shtydb.Liquid trol System llW verified b c performance of the follow erify.pump OPERABILXTT in ccordance'th Speci cation l.2.ht least once during each operating cycle: eck at t e se ting o thc s stem eli val s is 425 si gg s.l.1,8 anus nitiatc t e system, ept lo-s e valves Visua ly ver y flov pumpi g boroa solution through thc re rculati path d back o the S andby Li uid Con ol Sol'on T hftc umping ron olution, the sy cm hall be ushe vith dc'vater Verify mxnxmum BFH Unit 2 3.4/4.4-1 sR Z./.7.6 P0)t./tobe'.t.QT NA OF~S g gc ('eq4'o 3~~7 SEP 02 1888 gg 3.176 yump floM rate of 39 gpm against a system head of 1275 ysig pi demincra'=-er rom t+e St dby Li uk'ont.ol Test Tank.5+3~I'7'7 c~initiate one of the Standby Liquid Control System looys and nto e reactor vessel.is test checks lesion'he ch rge ass iatcd v h the teste loop, y er oyerati of the ives, and pump erabilit n car essa'selec d such hat he ag of cha ge Xn rvic 1 not eed f vc year from th manufacturer's assembly date<4 R./.7.7 net bo ex'v val~shal c teste n t e course of tvo operating cycles'FH Unit 2 3.4/4.4-2 Pg<<0'A~"}T l'.0.].5 0 s,f<J-,a~z.i.z FEB 2 8 tggg A>Ae.mod From and after the date that a redundant component is made or found to be inoperable, Specification 3.4.A.l shall be considered fulfilled and continued operation permitted provided that the component is returned to an operable condition within seven days.1.When component i found to b inoperable its red dant compo nt sha be emonstrate to be o erable imme iately d aily therea ter unti the operable ponent is repaired.t.Co 3 t.7 t all times+hen the Standby iquid Control System is required o be OPERABLE, the folloviag onditions shall be met: The olloving t ts s 1 be perfo ed to veri the availabi ty of the quid Control So ution: l&~17.9 I st3.l.7,3 1.At least 186 pounds Boron-10 must be stored ia the Standby Liquid Control Solution Tank and be available for injection. The sodium pentaborate solution concentration must be equal to or less than 9.2X by@eight.SR3 l ll SR~.l.l,+20 rg s.l.7.0 3~Volume: Check at least once per day.Sodium Pentaborate Concentration check at least once per month.Also check concentration within 24 hours anytime water or boron is added to the solution.Boron-10 Quantity: E OR V(r',Cp.S4.c, co~(e),~g; ~~pe~k>>r~,p <<444'>>are~IVII>>ply(I't iQUr(L>PiJ8 Fi)ure.g.l.7 I sa~I 79 4, At least once per month, calculate aad record the quantity of Boron-10 stored ia the Standby Liquid Control Solution Tank.Boron-10 Enrichment: At least once per 18 months and folloving each addition of boron to the Standby Liquid Contxol Solution Tank P~posecg~R p.l.'7.2 BHi Unit 2 3.4/4.4-3 AMENDMENT No.23 3 ,-,-~F

At Ogg 0 7 1594 5g 3./7.'f ae Calculate the enrich-ment within 24 hours.b.Verify by analysis within 30 days.4',D WR/75 The Standby Liquid Control System conditions must satisfy the following equation.rn g 1 (13 wt.)(86 gpm)(19.8 atom%)where, 5g R/7.S Verify that the equation given in Specification 3.4.D is satisfied at least once per month and within 24 hours anytime water or boron is added to the solution.C=sodium pentaborate solution concentration (weight percent)termine by the mo t rec nt pe ormsnce f the su eillhqce inst ction r uired by Specif ation 4.4.C.2.9~pump flow rate (gpm)termine y t c m st recount p formance f the s eil~ce ias ction r ired b Speci'cation 4.4.A.2.E~Boron-10 enrichment (atom percent Boron-10)Determined y the most repen erf o e of the+urveilhpce truction equired by Spec cation 4.4.C.4.)Pc.WceK 3 bC.If Specification 3.4.h through 3.4.D cannot be fset, make at least onc subsystem OPERABLE within 8 hours or the reactor shall be placed in a SHUTDOWN CONDITION with all OPERABLE control rods fully inserted within the followiag 12 hours.l.additkoaal c ance~reei red.BPÃUnit 2 3.4/4.4-4 NENDMNT NO.2"-9

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE

~~~et%~~f V f~(5C'~~~w~~~~~~~~~~~~~~~~~~~f~~~~~~~~4~~~~~~~~~~~I~~I~~'~~~I~~~~~~~~~~~~~~~~f~~~-WAa rVa.art't1%ItI I~~~~~~~~~~~~~I~~~~~~~~'~~~~~~~~~Il~~~~~~~I~~~~~~.~~~~~~~~~~~~~I~~~~~~~~~~~~~~~~~~~I~~~~~~~~~I~~~f~ 0 L (gi~~~~~o 5'g 3.I.7.g pump flov rate of 39 gpm against a syst ead o si umpi demi era ze v tcr om th St dby iquid ont 1 Tes Tank.+3'.l.7-7 c a initiate one of the Standby Liquid Control System loops and hai ra a to thc reactor vessel.s est ecks los n of hc ch ge a ocia d vi the te cd 1 p, pr er oper ion f the alve d p ac nt arg s s all b selcc d su t at hc age of ge i sc i shal not ceed iv years rom t e manufac urer'ss bl date 5/3.).'7,7 d.Both systems, u 0 v s a e tested in the course of tvo operating cycles.BFH Unit 3 3.4 4 4-2 AMENDMEWT NO.y y>PAGE~OF

5feCigcc son'3.l.7 FEB 2 8 L995 Qi 1.From and after the date that a redundant component is made or found to be inoperable, Specificati.on 3.4.A.1 shall be considered fulfi.lied and continued operati.on permitted provided that the component i.s returned to an operable condition wt.thin seven days.1.Whe a c pon t i fo d to be i per le, ts r duad t c pon t s 11 e dern nst ted o b opera le'di tel an dail the eaft r til the ino rab e co pon nt s rep ire 1.At least 186 pounds Boron-10 must bc stored in the Standby Liquid Control Solution Tank and bc availablc for injection. 2~5'R 8.1.7.3 The sodium pentaborate solution concentrati.on must be equal to or less than 9.2X by weight.At all times when the Standby Liquid Control System is required to be OPERABLE, the following conditi.ons shall be met: h foll wing t sts sha 1 be per armed to ver y thc avai abili of th Li.quid r Solu 5$3ii li I 1.Volume: Check at least once per day.sa a.~-~>2.Sodium Pentaborate Concentration check at least once per month.Also check concentrati.on within 24 hours anytime water or boron is added to the solution.SR 7 I-7-9 3.Boron-10 Quantity: yg,;Q~c~ceehahon+d wfarc nP boron I 5 So (4+ion~LO'i 44 t h Li fg pf Ci94rC 3 l 7 PDD Crgur~Z.I.7-/At least once pcr month, calculate and record the quantity of Boron-10 stored in the Standby Liquid Control Soluti.on Tank.SR Z.l-1-9 4.Boron-10 Enrichment: At least once per 18 months and following each addition of boron to the Standby Liquid Control Solution Tank Progress 5 3/72 BBl Unit 3 3.4 4.4-3 0 ~g 0 7 3994~53.i 7'I a.Calculate the enrich-ment within 24 hours.b.Verify by analysis within 30 days.)3.4.D R 3.lo"7i5 The Standby Liquid Control System conditions must satisfy the following equation.r 1.(13 wt.X)(86 gpm)(19.8 atom%)where, 4~4,D SR S.l.vs-Verify that the equation given in Specification '.4.D is satisfied at least once per month and within 24 hours anytime water or boron is added to the solution.C~sodium pentaborate solution concentration (weight percent)ete ned by he mo rece t rfo ce of he su eill ce i truct requ ed b e ificat on 4.4..2.Q~pump flow rate (gpm)De rmined by the most recent per rmance of tQ surveillance inst tion requir~by Specification 4.4 2.b.E~Boron-10 enrichment (atom percent Boron-10)I j}c4'un>8+C Dete ined by the mo recent perfo ce of the s illance instruct required by Specification 4.4.C.4.1.If Specification 3.4.A through 3.4.D cannot be met, make at least one subsystem OPERABLE within 8 hours or the reactor shall be placed in a SHUTDOWN CONDITION with all OPERABLE control rods fully inserted within the following 12 hours.1.N dditio al surv illance re uired.BFN Unit 3 3.4/.4-4 ep0vmr I e g 0 13USTIFICATION FOR CHANGES BFN ISTS 3.1.7-STANDBY LIQUID CONTROL SYSTEM ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 Surveillance Requirements for pump operability that are required by the Inservice Testing (IST)'rogram have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.TECHNICAL CHANGES-MORE RESTRICTIVE Ml Added Surveillance to verify the continuity of the explosive charge.The continuity check is intended to ensure proper operation will occur if required.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Verification of the relief valve's proper operation and setpoint is conducted in accordance with the plant's Inservice Test Program and the ASME code.LA2 The method of performing surveillance tests is relocated to plant procedures. LA3 Requirements on the replacement charges for explosive valves have been relocated to the Bases and plant administrative controls.BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.1.7-STANDBY LIQUID CONTROL SYSTEM TECHNICAL CHANGES-LESS RESTRICTIVE CONTINUED"Specific" Ll The CTS states applicability is at all times when fuel is in the vessel and the reactor is not in a shutdown condition with BFN TS 3.3.A.l satisfied. The proposed ISTS Specification does not require SLC System operability during Hot Shutdown, Cold Shutdown, or Refueling (Modes 3, 4,&5)since control rod withdrawal is limited and adequate SDM prevents criticality under these conditions. L2 Added the second part of SR 3.1.7.3, which provides the flexibility of allowing the concentration of boron in solution to be greater than 9.2%by weight as long as it is within the limits of proposed Figure 3.1.7-1 and the equation of SR 3.1.7.5 is met.Figure 3.1.7-1 has been added to allow this flexibility. This is acceptable since there is a 10'F thermal margin to unwanted precipitation of the sodium pentaborate. Per FSAR Chapter 3.8.3, the worst case sodium pentaborate solution concentration required to shutdown the reactor with sufficient margin to account for 0.05 M/k and Xenon poisoning effects is 9.2 weight percent.This corresponds to a 40'F saturation temperature. The worst case SLCS equipment area temperature is not predicted to fall below 50'F.SR 3.1.7.3 must be performed within 8 hour s of discovery that the concentration is)9.2 weight percent and every 12 hours thereafter until the concentration is verified a 9.2 weight percent.This Frequency is appropriate under these conditions taking into consideration the SLC System design capability still exists for vessel injection and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.L3 Deleted BFN TS 4.4.B.1, which requires that when a component is found inoperable, its redundant component be demonstrated operable immediately and daily thereafter until the inoperable component is repaired.This requirement is deleted for several reasons.Increased testing has not been shown to demonstrate operability any better than testing at the normal SR test interval.In many cases, increased testing adds to the failure rates of components by increasing wear and tear.Common mode failure analysis in conjunction with loss of function analyses provide adequate assurance of redundant system operability. Loss of function determination program controls are provided by BFN ISTS 5.5.11.BFN-UNITS 1, 2,&3 Revision 0 pi ci UNIT 1 CURRENT TECHNICAL SP ECIF ICATION MARKUP

SAC>g;cq,/on P, I, P QcC 0 7 1994.3.E I-v If Specifications 3.3.C and.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours.4.3.E Surveillance requirements are g as specified in 4.3.C and.D ve S<<Tugtj hea+'on 4 (Chalets BPw)S75 3.l.w 4 3,l.p e OPERA hat the reactor protection system is re uired to be OPERABL xce cx wed in 3.3.F.2.r~cd'loke& 5~$oflkHou Table 1.The scram discharge vol g.~3,I,Q drain and vent valves hall I<R 3,1.P,)l.a.SN 3 1.S.)hJ+W The scram discharge volume drain and vent valves shall be verified o I and monthly thereafter. The valves may e closed intermittently for esting o o exceed any 2 hour eriod ing operation. P~r]'o SX a,l.S.2.The scram discharge volume drain and vent valves shall be demons tratcd OPERABLE in accordance with Specification 1.0.MM.2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided thc redundant drain or vent valve is OPERABLE 2~Wh i.t i dete ined t any SDV dr in or v t va e is opcrab e, t e red dant rain o v t v ve sh 11 be cmon rated OPERAB imme ately d wee ly thercaftcr Ron C Ckf'ioH 3.f redundant dra n or vent 5 valves become inoperabl e'xn HOT SRAiHBRY CONDITION within 44-hours.12-No diti al surve'lian e req ired.5LWds~n Pcnp$4~R Z,l 93 BFN Unit 1 3.3/4.3<<12 hMEVWENr NO.g g>PAGE UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

gKC 07 1994 3.3.E.4.3.E.If Specifications 3.3.C and.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours.)gl P~~pceCh Nrr4 a+she P ACTloPU 7qklg l.The scram discharge volume drain and vent valves shall CAJ Iran e OPE any tame that the reactor protection system is OPERABLE cept as ecx ted in 3.3.F.Z.Sg 3.).E.l.a.The scram discharge volume drain and vent valves shall be verified o Mg~2 mon thercaftcr. The valves may e closed intermittently for Sg, s.t.a.l Wo4e testing no aqccce hour in y 24-haur per duri erat on.Surveillance requirements are as specified in 4.3.C and.D above See.V&4ic,akisn 4vi-0"4~)<Ar OFT l~3,t,g p g J.f pg,1is<A 2.In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.5g 3rlrsr~b 20 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.%f.When t is determ ne tha any S drain or t valve is operab, the redund t drain o vent lve shal bc d s tra ted ERABLE iately d weekly thereafter. L<3.If redundant dra or vent Acr(ad CONDITION within C PC hours.s~W" A~s cill re rcd.pro("~sg 3.L.Q.Q BFN Unit 2 3.3/4.3-12 NENggg gg, g 29 PAGE~GP~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 0 PeCon.I.8 0 tS>4 V 4.3.E.If Specifications 3.3.C and 3.3.D above cannot be met, an orderly shutdown shall be initiated and the reactor.shall be in the SHUTDOWN CONDITION within 24 hours.Surveillance requirements are as specified in 4.3.C and 4.3.D above.z ac Tush C<cohen&r Ch4~Qc BPpl XSTS Sil2+3.I-'I Pr up5cd Lco 3.1 Q mi No>MS~f og 1 Sod'Able The scram discharge volume drain and vent valves shall e OP any tame that the reactor protection system is re'd to b OPERABL as ed in 3.3.F.2.sR'B.t.8.l SR Z,~.e,1 hru The scram discharge volume drain and vent valves shall be verified o en IQ mont y thereafter. The valves may be closed intermittently for testing n to e ur i any-h r per d d ing o rat 5/3, j,g,?.l.b.Thc scram discharge volume drain and vent valves shall be 3 demonstrated OPERhBLE in accordance with Specification 1.0.%4.2.In thc event any SDV drain or vent valve becomes inoperable, REhCTOR POWER OPERhTION may continue provided the redundant drain or vent valve is OPERhBLE.LI 20 When t is determined tha any d in or v val e is per le, e r dun t ain v t v ve hall be ra d OPERhB diately and weekly ereaftcr.3.If redundant drain or vent valves bccomc ino rable B r ac or e xn y'o t4 CONDITION within hour~3.No ad ti 1 urvei ance quired hh2 SAKES~<BFN Unit 3 3.3/4.3-12 NENDMNT NQ.y 86 PAGE+OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.8-SDV VENT AND DRAIN VALVES ADM IN I STRAT I VE CHANGES Al Reformatting and renumbering is in accordance with the BWR/4 Standard Technical Specifications (STS), NUREG-1433. As a result, the Technical Specifications (TS)should be more readily readable, and therefore understandable, by plant operators as well as other users.During the reformatting and renumbering of the improved Technical Specifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. A2~-CTS Surveillance Requirement 4.3.F.l.a requires that the SDV drain and vent valves be verified open PRIOR TO STARTUP.These words are unnecessary and were deleted to make the BFN ISTS SR 3.1.8.1 consistent with the BWR Standard Technical Specifications, NUREG-1433. Proposed SR 3.1.8.1 requires the valves to be verified open when they are required to be operable in Modes 1 and 2.Proposed SR 3.0.4 does not allow entry into a Mode unless the SRs have been met within their specified frequency. Therefore, this SR is required to be met prior to entry into Mode 2 or"prior to startup." Since the intent of the SR is not changed, the deletion of these words are considered administrative. CTS 4.3.F.l.b requires the SDV drain and vent valves to be demonstrated OPERABLE in accordance with Specification 1.0.MM, which is the Surveillance Requirements for ASHE Section XI Pump and Valve Program.This program provides equivalent testing requirements, with respect to valve cycling not closure times, to proposed SR 3.1.8.2, which requires each SDV vent and drain valve to be cycled fully closed and open every 92 days.Therefore, the proposed change is considered administrative. A4 Deleted CTS 4.3.F.3, which states no additional surveillance required, to make the BFN ISTS consistent with NUREG-1433. It is unneces'sary to specify that no additional surveillance is required-omission of this statement would serve the same purpose.Therefore, the proposed change is considered administrative. A5 The Note in proposed SR 3.1.8.1 provides an allowance that does not require the surveillance to be met on SDV vent and drain valves that are closed during the performance of SR 3.1.8.2, which requires valves to be cycled fully closed and open every 92 days.CTS allow the valves to be closed intermittently for testing but this is not allowed to exceed 1 hour in any 24-hour period during operation. Since each SDV vent and drain valve is required to close in x 60 seconds per proposed SR 3.1.8.3, the current 1 hour allowance for the valves to be closed for BFN-UNITS 1, 2, 5.3 Revision 0 PAGE~OF~

JUSTIFICATION FOR CHANGES BFN ISTS 3.1.8-SDV VENT AND DRAIN VALVES ADH IN I STRATI VE CHANGES CONTINUED testing in any 24-hour period will not be exceeded when cycling the valves to the fully closed and fully open position.Since the intent is the same (i.e., to allow the SDV vent and drain valves to be cycled during reactor operations), the proposed change is considered to be administrative. TECHNICAL CHANGES-NORE RESTRICTIVE Ml CTS 3.3.F allows unlimited continued operation when any SDV drain and vent valve becomes inoperable provided that the redundant drain or vent valve is demonstrated OPERABLE immediately and weekly thereafter. Proposed Action A is more restrictive since it allows continued operation for 7 days.At that time if the valve has not been restored to'PERABLE status, the reactor must be placed in MODE 3 within 12 hours.Proposed Action C requires the plant to be in MODE 3 in 12 hours while CTS 3.3.F.3 requires the plant to be in HOT STANDBY CONDITION (equivalent to MODE 2 at a 1%RTP)within 24 hours of redundant drain or vent valves becoming inoperable. Proposed Action C is more restrictive since it does not allow as much time to change modes and requires the reactor to be placed in MODE 3 versus HOT STANDBY (equivalent to MODE 2 at c 1%RTP of proposed BFN ISTS).M3 Added SR 3.1.8.3, which requires that an integrated test of the SDV vent and drain valves be performed on an 18 month frequency to verify total system performance. After the receipt of a simulated or actual scram and subsequent scram reset signal, the closure and subsequent opening of the SDV vent and drain valves, respectively, are verified.The closure time of 60 seconds is acceptable based on the bounding leakage for release of reactor coolant outside containment. The LOGIC SYSTEM FUNCTIONAL TEST in Proposed LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function.BFN-UNITS 1, 2, L 3 Revision 0 PAGE~pe~

13USTIFICATION FOR CHANGES BFN ISTS 3.1.8<<SDV VENT AND DRAIN VALVES TECHNICAL CHANGES-LESS RESTRICTIVE Ll Added a proposed Note (" Separate Condition entry is allowed for each SDV vent and drain line")at the start of the ACTIONS Table to provide more explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3-"Completion Times," this Note provides direction consistent with the intent of the proposed Actions for inoperable SDV vent and drain valves.Each.SDV line is intended to be allowed a specified period of time to confirm it isolated or is capable of isolation, and to restore the complete function of the line.Current TS 3.3.F.3 requires the reactor to be in Hot Standby Condition within 24 hours if both valves are inoperable in one or more SDV vent or drain lines.Proposed Action B allows 8 hours to isolate the line(s).Both valves must be restored to operable status within 7 days per Action A.Recognizing that the SDV vent and drain valves are normally open to prevent accumulation of water in the SDV from leakage, a Note has been added to Required Action B.1 (which requires isolation of the line), allowing periodic opening of the affected line for draining and venting the SDV.This may be necessary due to CRD seal leakage in order to avoid automatic reactor scrams on high level in the SDV.These extended times, and the option to administratively un-isolate a SDV line isolated by a Required Action, are consistent with the BWR Standard Technical Specifications, NUREG 1433.These increased allowances are deemed not to substantially increase the risk of a scram with an additional failure that could allow the SDV to remain un-isolated; nor to substantially increase the risk of the SDV failing to accept the control rod drive water displaced during a scram.L2 CTS 3.F.1 requires the SDV drain and vent valves to be OPERABLE any time that the reactor protection system (RPS)is required to be OPERABLE.Proposed BFN ISTS 3.1.8 requires the SDV vent and drain valves to be OPERABLE in Modes 1 and 2.Currently, portions of the RPS are required to be OPERABLE during other MODES, as described in BFN TS Table 3.1.A,'herefore, the proposed Specification is considered less restrictive. The proposed Specification applicability is based on when a full scram may be required.In MODES 3 and 4, control rods are only allowed to be withdrawn under proposed Special Operations LCO 3.10.3 and 3.10.4, which provide adequate controls to ensure that only a single control rod can be withdrawn. Also, during MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. The SDV vent and drain valves need not be OPERABLE in these MODES since the reactor is BFN-UNITS 1, 2, 5.3 Revision 0 PAGE 3 oj= JUSTIFICATION FOR CHANGES BFN ISTS 3.1.8-SDV VENT AND DRAIN VALVES TECHNICAL CHANGES-LESS RESTRICTIVE CONTINUED subcritical, only one rod may be withdrawn, and the SDV is adequate to contain the water from the single rod scram even if isolated.L3 Deleted BFN TS 4.3.F.2, which requires that when a component is found inoperable, its redundant component be demonstrated operable immediately and daily thereafter until the inoperable component'is repaired.This requirement is deleted for several reasons.Increased testing has not been shown to demonstrate operability any better than testing at the normal SR test interval.In many cases, increased testing adds to the failure rates of components by increasing wear and tear.Common mode failure analysis in conjunction with loss of function analyses provide adequate assurance of redundant system operability. Loss of function determination program controls are provided by BFN ISTS 5.5.11.BFN-UNITS I, 2, 5 3 Revision 0

UNIT 1 CURRENT TECHNICAL SP ECIFI CATION MARKUP PAGE~OF -5 ggllC'+AD 3.4cT4++)I:c L.P.'ing steady-state povcr o cration c vcragc anar Linear Heat Generatfon Rate (hPLHGR)of any fuel assembly at any axial location shall not exceed the appropriate kPLHGR limit provided in thc CORE OPERhTI?lC ZMZTS REPORT.If at any tfae during steady state operation ft is determined by normal surveillance that the limitiag val is being exceeded ction ahull be atc~minutes tW est o eration to Vf the P thfa rescrfbed limits If the kPLHGR fs not rcturncd to vfthin the prescribed lfaits vithfn tvo (2)hours, the reactor shall be brought to XhPLKBl gg g.a I.I Sie kPLHQR shall bs checked daily during reactor operation at g 25K rated thermal pover.Q Ls'F~Jcq p ging ac shall con~ue until ctor operat is vithin the prcscri limits.Bur%a steady-state pover operation, the linear heat generation rate (LHCR)of aay rod fn any fuel asseably at any axial location shall not exceed the appropriate LHCR liait prorfded in the CORE OPERATIC LINZT5 REPORT o The LHCR shall bc checked daily during reactor operation at g 25K rated thermal pover.<<<<~4ÃcgI. ISTIC).g g BHf Uait 1 3 5/4.5 18 AMENOMHff N.I 9 7 UNIT2 CURRENT TECHNICAL SPECIFICATION MARKUP 0 0 M 20 1993 LCO S,Q I Apt Lck'.I, ring steady-state pover operation e vcragc anar Linear Heat Generation Rate (APLHQR)of any fuel assembly at any axial location shall not exceed thc appropriate hPLHGR limit provided in the CORE OPERATIHQ LIMITS REPORT.I f at any time during operation it is determined by normal surveillance that the limiting value for kPLHQR is bei cxcceded, s ge LL n at vithin 5 minu to esto operation t vithin the rcscrib limi.I the hPLHQR is not returned to vithin the prescribed limits vithin tvo (2)hours, the reactor shall be brought to urvc ance corr capon ac oh shall cuutfuue uutilll react operation is vithin the prcscri ed limits 4..~v G SR Z*l I Xhe aPLHGR shall be cheeked daily during reactor operation at g 25X rated thermal povcr.tb~tf4><~4 frep~C~r+e C XR 2 g.I.J<zsY a7-t.Wi4h:~q4,~~r.During sCeady sCaCe povcr operation) the linear heat generation rate (LHQR)of any rod ia any fuel assembly at any axial location shall not exceed the appropriate LHQR liakt provided in the CORE OS'~?I LIMITS REPORT.If at any oCaring operacion ic is dloilatacd by normal surveillance that the liaitiag value for LHQR is being exceeded, action ahall be'initiated vithin 15 ainutcs to restore operation to vithia the prescribed limits.If the LHQR is not returned to vithin the prescribed limits vithin tvo (2)hours, the reactor shall bc brought to the COLD SHUTDOWN COIDITIOE vithia 36 hours.Surveillance and The LHQR ahall be checked daily duriag reactor fuel operatioa at g 25K rated thermal pover.Scc.a~it'Aa4'o <r C4 Ivy-Br~>>7 s Z.2.3 B2$Unit 2 3.5/4.5-18 0 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE/OF l.co S.Q./nieN A ring steady-state povcr o eration the Average lanar 1'c L.>'rve ce corrcspo act shall cont anti~actor opcrat ls vlthin the s b ts+inear Heat Generation Rate (APLHGR)of any fuel assembly at any axiDzJ.location shall not exceed the appropriate APLHGR limit provided in the CORE OPERATIC ZMZTS REPORT.If at azqr time during operation lt is determined by normal surveillance that the limiting value for APLHGR ls being exceeded, ct on c tlatcd 5 minutes to store perat vith thc imit If the APLHGR ls not returned to vithin thc prescribed limits vithln tvo (2)hours, th reactor shall be brought to c f I sR 3.B.I.e APLHGR sh u.be chc~cd I daily daring reactor operation at g 25K rated thermal pover.PJd pn<r F I cp vt<<J o C Jg, 3.2.l.I g.TP Q ha~or Daring steady-state pover operation, the linear heat generation rate (LHCR)of any rod in any fuel assesbly at any axial location shall not exceed the appropriate LHCR limit provided ln the CORE OHRATIIC LASTS REPORT.The LHCR shall be checked daily daring reactor operation at 2, 25K rated thermal pover.If at any time daring operation it is determined by normal sarvelllance that tha limiting value for LBCR ls behlg.exceeded, action shall be lnitfatsd vlthin 15 minutes to restore operation to vithln the prescribed limits.If the LHCR is not retarned to vithin the prescribed limits vithln tvo (2)hours, the reactor shall be brought to the COLD SHUTDOWNÃCONDITION vlthln 36 hours.Surveillance and corresponding action shall continue until reactor operation is vlthln the prescribed limits.Sg~~~4Pica4o 4~r C4-g~~<SF's7.s z.z.3 BHf Unit 3 3.5/4.5-18 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.2.1-AVERAGE PLANAR LINEAR HEAT GENERATION RATE ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 The Appl.icability has been changed from"steady-state power operation" to"Thermal Power a 25%RTP." This change is considered administrative in nature since, based on a CTS surveillance frequency of daily during reactor operation at a 25%rated thermal power, the intent of CTS is the same as the proposed ISTS specification. A3 The requirement to continue the surveillance when the limits are not met has been deleted since the total allowed completion time for restoring the limit or placing the plant in a condition outside the Applicability is 6 hours.Since the 6 hour time frame is less than the Surveillance Frequency of 24 hours, the surveillance would not be required to be performed again while the plant was in the action.The requirement to continue to comply with actions until the limits are met has been moved and is now addressed by LCO 3.0.2, which clarifies that if an LCO is met or is no longer applicable prior to the.expiration of the specified Completion Time(s),,completion of the Required Action(s)is not required, unless otherwise stated.As a result, these changes are~considered administrative in nature.TECHNICAL CHANGE-MORE RESTRICTIVE Ml A new Frequency has been added to require verification of APLHGR within 12 hours of reaching or exceeding 25%RTP.This is an additional restriction on plant operation. CTS would allow up to 24 hours after reaching 25%RTP to perform the test. JUSTIFICATION FOR CHANGES BFN ISTS 3.2.1-AVERAGE PLANAR LINEAR HEAT GENERATION RATE H2 The requirement to place the plant in a COLD SHUTDOWN CONDITION within 36 hours when the limit is not restored within the required completion time has been revised to reflect placing the plant in a non-applicable condition. Current Specification 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in cold shutdown is not applicable after thermal power is reduced below 25%RTP.The revised Action requires plant power to be reduced to<25%RTP (outside the applicable condition) within 4 hours.The cur rent action allows 36 hours to place the plant in a non-applicable condition. As such, this is an additional restriction on plant operation. TECHNICAL CHANGES-LESS RESTRICTIVE"Specific" Ll The requirement to initiate action within 15 minutes to restore the limit is relaxed and relocated to the Bases in the form of a discussion that"prompt action" should be taken to restore the parameter to within limits.Immediate action may not always be the conservative method to assure safety.The 2 hour completion time for restoration of the limit allows appropriate actions to be evaluated by the operator and completed in a timely manner.PAGEOF

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 S oci lif%4lo an s.Z.Z.FEB 8 41995~5.J 4.5~3.5.J (Cent'd)If at axe time during steady-state o eration it is determined by P normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to vithin the prescribed limits.If the LHGR ia not rcturncd to vithin thc prescribed limits vithin tvo (2)hours, the reactor shall bc brought to the COLD SHOTDOMH COHDITIOH vithin 36 hours.Surveillance and corresponding action shall continue until reactor operation ia vithin the reacribcd limits.AC Z~l;C;~l('~ 4r C4,)~gfr4 ISrS Z.2 3 d p~pr~ISf F~.sC$8@.2.2.(L.CO 3.2.2.+l~l,q)i'm>oQ A 4g Q ARRl Thc minimum critical pover ratio (NCPR)ahall be equal to or greater than the operating, limit MCPR (OLMCPR)aa provided in the CORE OPERATIHC LIMITS REPORT.If at any time during steady-stat it ia determined by normal surveillance that the limiting value for MCPR ia bc ezceeded, ction shall e ti within l~qpt~o eatore o ion to vithin rsacribed limits e at04df-stats MCPR is not returned to vithin the praacribad limits vithin tvo (2), the reacto 11 bc brought to urvc ance correa ng act ahall c ue anti.react operation vlthin reacribcd limits.g 2.2.1.MCPR shall be cheeked daily daring reactor pover operation at g 25X rated thermal povcr o ov ng c pover vel or atribu that v d cause eration ith a LINITIIC COHTROL RO PLTXZRK SCS.>2~2.The NCPR limit at rated flov and rated povcr shall be detcrmincd aa provided in the CORE OPSRhTIHC LIMITS REPORT using I a.aa defined in the CORE 0 LIMITS REPORT prior o initial scram time mc cnts the cycle, par d in accordance vith 8 if ication 43 C.l BHt Unit 1 8 2,$'Kd W~F)Qo~4/Le<<$3 5/4.5-19 iNENDMBlT NO.2 1 6

FES 2 4>ggg g.AL b.as defined in the CORE OPERA IHG LIMITS PORT foll ing the c elusion of ecr time surveil ce test r uired by Spec-fications 4.3.C.l and L.L Sg 3.Z.~~The determination of the limit mast be completed vithin 72 hours of each scram-time surveillance required by Specification 4.3.C.1.Whenever the core thermal pover is g 25K of rated, the ratio of FRP/CMFLPD shall be g 1.0, or the kPRN scram setpoint equation listed in Section 2.l.k and the hPRN rod block setpoint equation listed in the CORE OPERiTISC LQGTS REPORT shall be amltiplied by FRP/CÃFLPD. FRP/CNFLPD shall be determined daily vhen the reactor is g 25K of rated thermal pover.2.lthen it is determined that 3.5.L.I is not beiag met, 6 hours is alloved to correct the conditiono 3~.If 3.S.L.1 and 3.5.L.2 cannot be mat, the reactor pover shall be reduced to S 25K of rated thermal pover vithin 4 hours.BF5 Unit 1 3.5/4.5-20 eeMENr Na 2 Z6 PAGE~pF~ o 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 0' BVSdP~4i~~'0~c i!li'~~~~81rIOOBIa~'fir rvralif B4li%lit%+i'f56 ~~~~~~~4~~~~~~~'~~~~~~~~~~~~I I tip III~l I I I~~I~~~~~~~~~I~~~~~~~'~~'~g~~~~~~~~~~~~~~~'~~~~I~I~~~I~~~~~~~~~~~~~~~~~~~~~~~~~~'I~~~~~~~~I~~~~~~~~~'~~~~~~~~~~AO~~~I sTo~~~~~~~~~~~~I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I~~~~~~~~~~~~ UNIT 3 CURRENT TECHNICAL SP ECIF I CATION MARKUP PAGE 'x EB 2 4 1995 LCo X2.2.~p~'cd.I, An'(owl A Thc mfnfmum critical povcr ratio (MCPR)ahall be equal to or greater than the operating limit MCPR (OLMCPR)as provided fn the CORE OPEBLTIHG LIMITS REPORT.If at auy t ur stead stat 0 it ia determined by normal surveillance that the limiting value for MCPR fs being exceeded, C OIl a e 5 ea to reato operation to rcacribcd limit.If thc steady-state MCPR ia not returned to vithin thc prescribed imits vithln tvo (2)hours, the reactor shall bc brought to thc~ours, urve ance orrea ing acth@shall~oat fnue untQ reactor operation la vkthfn thc prMcrfbed limits~2S 7',vp+2 Vip'4gg~g S'R MCPR 1fait at rated flov and rated yover shall be determined as provided in the CORE OPERATIEQ LIMITS REPORT 44l as fined in thc CORE 0 KG LIMITS REPORT yrio to initfal scram t measurements for the cle, yerfo ed in ccordance th Specifics on 4.3.C.l.as d fined in c CORE OPERAT L REPORT folio th conclusion of each acr tfmc surveillance teat required by Specificationa 4.3.C.l 4 3.C.2~sa dctcrmfnition of the 1iaft must be completed vithf11 72 hours of each scram-time surveillance requfred by Specification 4.3 C.5'g B.z 2.I during reactor yovcr operation at g 25X rated thermal pov o ov any e n p ver level o distribution vould caus operation vith LIMITIEQ COSTROL RO PATTERE BlÃUnit 3 3.5/4.5-19 NENDMEHT NO.I 9 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.2.2-HININUM CRITICAL POWER RATIO ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 The Applicability has been changed from"steady-state operation" to"Thermal Power w 25%RTP." This change is considered administrative in nature since, based on a CTS surveillance frequency of daily during reactor operation at a 25%rated thermal power, the intent of CTS is the same as the proposed ISTS specification. A3 The requirement to continue the surveillance when the limits are not met has been deleted since the total allowed completion time for restoring the limit or placing the plant in a condition outside the Applicability is 6 hours.Since the 6 hour time frame is less than the Surveillance Frequency of 24 hours, the surveillance would not be required to be performed again while the plant was in the action.The requirement to continue to comply with actions until the limits are met has been moved and is now addressed by LCO 3.0.2, which clarifies that if an LCO is met or is no longer applicable prior to the expiration of the specified Completion Time(s), completion of the Required Action(s)is not required, unless otherwise stated.As a result, these changes are considered administrative in nature.TECHNICAL CHANGE-MORE RESTRICTIVE Hl A new Frequency has been added to require verification of MCPR within 12 hours of reaching or exceeding 25%RTP.This is an additional restriction on plant operation. CTS would allow up to 24 hours after reaching 25%RTP to perform the test.BFN-UNITS 1, 2, 8L 3 Revision 0 PAGE~gp

JUSTIFICATION FOR CHANGES BFN ISTS 3.2.2-MINIMUM CRITICAL POWER RATIO H2 The requirement to place the plant in a COLD SHUTDOWN CONDITION within 36 hours when the limit is not restored within the required completion time has been revised to reflect placing the plant in a non-applicable condition. Current Specification 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in cold shutdown is not applicable after thermal power is reduced below 25%RTP.The revised Action requires plant power to be reduced to<25%RTP (outside the applicable condition) within 4 hours.The current action allows 36 hours to place the plant in a non-applicable condition. As such, this is an additional restriction on plant operation. TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LA1 The method used to determine r is moved to the Bases in the form of a discussion (describing the ways to compute r).This information is also contained in the Core Operating Limits Report (COLR).The proposed change does not change the intent of CTS."Specific" Ll The requirement to initiate action within 15 minutes to restore the limit is relaxed and relocated to the Bases in the form of a discussion that"prompt action" should be taken to restore the parameter to within limits.Immediate action may not always be the conservative method to assure safety.The 2 hour completion time for restoration of the limit allows appropriate actions to be evaluated by the operator and completed in a timely manner.L2 Since a limiting control rod pattern is currently defined as operating on a power distribution limit such as MCPR, the condition is extremely unlikely and the surveillance would almost never be required.In the CTS, determination that the plant is operating with a limiting control rod pattern would be found during performance of the daily SRs for thermal limits.If operating with a thermal limit in excess of CTS limits, proper actions are required to restore the plant to within limits.To ensure that the plant is restored to within limits, the SRs must be performed anyway, thus the additional SR frequency during limiting control rod pattern is not necessary. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE~OF ah

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP

5'peeila;J' <2.3 Al HN aos 3~501 4.5.I During steady-state povcr operation, the Average Planar Linear Heat Generation Rate (kPLHGR)of any fuel aaaeably at any axial location shall not exceed the appropriate kPLHGR limit provided in the CORE OPZRATIWC LIMITS REPORT.If at any time during steady state operation it ia determined by normal surveillance that the limiting value for APLHCR ia befng exceeded, action shall be initiated within 15 minutes to restore operation to vfthin the prescribed limits.If the APLHGR ia not returned to vithin the prescribed limits vithin tvo (2)hours, the reactor shall be brought to the COLD SHUTDOWNÃCOHDITIOS vithin 36 hours.Surveillance and corresponding action shall continue until reactor oyeration ia vithin the prescribed The APLHGR shall be checked daily during reactor operation at 2 25'X rated t over ggg~~ks4iQ~4ev g<QPN lsrs zz/LCO 3.2.3 Pppli~~y f1lg ate a the linear heat generation rate (LHCR)of any rod in any fuel assembly at any axial location shall not exceed the apyroyriate LHCR limit provided in the CORE OPERATISC LIMITS REPORT.QJHaRl gg 9 2m 3 LHCR shall be checked daily during reactor operation at g 25%rated thermal power.I"~~p p~4ey~~oI SK Rz..3.(BHf aait 1 3 5/4 5-18 AMENOMEHT N.g 9 7 0 .r Spscg~<Wiry<g Q3 FEB 8 41995 AT>eeJ Parian'If at any time during steady-state operation it is determined by normal surveillance that the limiting value for LHGR ia being exceeded ctlon shall e t cd vithinM~inu~~o eatorc ration to vithin the rescrib thc LHGR ia not returned to vlthin the prescribed limits vithin tvo (2)hours, the reactor shall be brought to re u ance c 0 ng action hall con ue unti reactor ration ia ithin th reacribed limits.<zs9,'.~~S~Zg4'Ace/g~g.r P,)i~4r SF'sr'z.z. 4.5.K The minimum critical pover ratio (MCPR)shall bc equal to or greater than the operating limit MCPR (OLMCPR)aa provided in thc CORE OPERATIHG LIMITS REPORT.If at any time during steady-state operation it is determined by normal surveillance that the limiting value for MCPR is being ezceeded, action shall bc initiated vithin 15 minutes to restore operation to vithin the prescribed limits.If the steady state MCPR ia not returned to vithin the prescribed limits vithin tvo (2)hours, the reactor shall be brought to the COLD SHUTDOWN COHDITIOI vithin 36 hours, surveillance and corresponding action shall continue until reactor operation is vithin the prescribed limits.l.MCPR shall bc checked daily during reactor pover operation at g 25Z rated thermal pover and folloving any change in pover level or distribution that vould cause operation vith a LIMITISG CONTROL ROD PATTZRE.2.The MCPR limit at rated flov and rated povcr shall be dctcrmined as provided in the CORE OPERLTIHG LIMITS REPORT using: a..as defined in the CORE OPERLTIHG LIMITS REPORT prior to initial scram time measurements for thc cycle, performed in accordance vith Specification BFÃUnit 1 3 5/4 5-19 memrr~o.a z 6 PAGE~OF~ 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 0 S cia'Agio~R 2.Q M 20 1993 3.5.I During steady-state pover operation, the hverage Planar Linear Heat Generation Rate (hPLHGR)of any fuel assembly at any axial location shall not exceed thc appropriate hPLHGR limit provided in the CORE OPERhTIHQ LIMITS REPORT.If at any tiae during operation it is detcrmincd by normal surveillance that the limiting value for hPLHGR is being exceeded, action shall bc initiated vithin 15 minutes to restore operation to vithin the prescribed liaits.If thc hPLHGR is not returned to vithin the prescribed limits vithin tvo (2)hours, the reactor shall be brought to the COLD SHOTDOWH COSDITIOlf vithin 36 hours.'urveillance and corresponding action shall continue until reactor operation is vithin the prescribed limits.l The hPLHGR shall be checked daily during reactor operation at g 25K rated thermal pover.peg JNg$i Ci(Jld H~+<C l4%$P Ar 8fhl<stan~2 I L.Cd 3.?.g gyve the linear heit generat on rate (LHQR)of any rod in any fuel asseably at any axial location shall not exceed the appropriate LHQR liait provided in the CORE OPIRiXIM LIMITS REPORT.If at any isE during operation it is deCazained by normal surveillance that the liaiting value for LHGR is being exceeded, ct on c i~ainutes cstoOQpcration to rescribed If the LHGR is not returned to vithin the rescribed liaits vithin tvo (2)hours the reactor shall bc brought to P ppl;a,L;t' During, steady-state pover operation QJKRl daily during reactor fuel operation it g 25Z rated thermal povcr.P 4d p~pgW Is I-F y r~cy dg S/i!3.g,.3.t g gg Vo RTP ivlg.gf 4@~BFH Unit 2 3.5/4.5>>18 AMENDMENT No.2 I 4 l AVER" I-~ 0 0 DEC 07 594 3.S.K corr ponding action s 1 contin til reactor op tion ia vithin c yreacribed limits.The minimum critical yover ratio (NCPR)shall be equal to or, greater than the operating limit NCPR (OLNCPR)as provfdcd in the CORE OPERhTISQ LINITS REPORT.If at any time during steady-state operation it is deterained by normal surrcfllancc that thc limitfng value for NCPR is being exceeded, action shall be initiated vithin 15 minutes to restore operation to vithin thc prescribed liafts.If the steady-state NCPR ia not returned to vithin the prescribed 1faita vithin tvo (2)hours, the reactor shall bc brought to the COLD SHUTDORf CO%)ITIOUS vithin 36 hours, surveillance and corrcayondfng action shall continue until reactor oyeration ia vithin the prescribed limits.5'~girSA'~iCrAao~ 4<C 74 SPY%%d gpnJ is V'S 7.z.z.1.NCPR shall be checked daily during reactor pover operation at g 25%rated thermal povcr and folloving any change in pover level or distribution that vould cause oyeration vith a LINZTIRC COITROL ROD PLTTEW.2.The NCPR limit at rated flov and rated yover shall be determined as provided in the CORE OPERATIC LINZTS REPORT using ae 4 aa dcfiILed fIL thc CORE OPERLTIÃC LINZTS REPORT prior to initial scram ciRO mcasurcmcnta for thc cycle, performed fIL accordaILCC vith Specification 4.3.C.1.b.4 aa dcffncd in thc CORE OPERATIC LINZTS REPORT follovtng the conclusion of each seri-tfae surveillance teat required by Sycciffcationa 4.3.C.1 and 4.3.C.2.The deterainatfon of thc 1iait aust bc coepleted vithin 72 hours of each acraa-tfac surveillance requfrcd by Specification 4.3.C.BHf Unit 2 3.5/4.5-19 AMENDMBIT 10.2 2 9<<7 pAGE~'~'- 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF ~r a'~'Ir T~TTltr t\r~LVI t1'IVa r'Vrv E~JJLtvvv a~~%r%%N.Il~Tv%'rl a.rear rv:~aa.-r~ar var~~"-~~~~~4~~~I~~~~~~~~~~~~~~~~~~4~~: I~~~~~~~~~~~~'I~L'~~'~~~~~~~~~r~~I~~~~~~~~I~~~~~~~~~~~~~~~~~I~~'~I I~~jl~~~~~~~~~~~~~~~tI~~~~~~~~~~~~~~5@ma~~~~~~~~~~~~~~V~~~~g~4'~~~~~~A,~~~~~~~~~~~~~L~~~~~~~~~~'~~.~aei~~...~~~I~~~~~~~~~~~~\~~~~~~~~~~~~~~~~~~~~~~~~I~~~~~~~~~~~~'I~I~~~~~~~ JUSTIFICATION FOR CHANGES BFN ISTS 3.2.3-LINEAR HEAT GENERATION RATE ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 The Applicability has been changed from"steady-state power operation" to"Thermal Power a 25%RTP." This change is considered administrative in nature since, based on a CTS surveillance frequency of daily during reactor operation at a 25%rated thermal power, the intent of CTS is the same as the proposed ISTS specification. A3 The requirement to continue the surveillance when the limits are not met has been deleted since the total allowed completion time for restoring the limit or placing the plant in a condition outside the Applicability is 6 hours.Since the 6 hour time frame is less than the Surveillance Frequency of 24 hours, the surveillance would not be required to be performed again while the plant was in the action.The requirement to continue to comply with actions until the limits are met has been moved and is now addressed by LCO 3.0.2, which clarifies that if an LCO is met or is no longer applicable prior to the expiration of the specified Completion Time(s), completion of the Required Action(s)is not required, unless otherwise stated.As a result, these changes are considered administrative in nature.TECHNICAL CHANGE-NORE RESTRICTIVE Hl A new Frequency has been added to require verification of LHGR within 12 hours of reaching or exceeding 25%RTP.This is an additional restriction on plant operation. CTS would allow up to 24 hours after reaching 25%RTP to perform the test.BFN-UNITS I, 2,&3 Revision 0 PAGE~OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.2.3-LINEAR HEAT GENERATION RATE H2 The requirement to place the plant in a COLD SHUTDOWN CONDITION within 36 hours when the limit is not restored within the required completion time has been revised to reflect placing the plant in a non-applicable condition. Current Specification 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in cold shutdown is not applicable after thermal power is re'duced below 25%RTP.The revised Action requires plant power to be reduced to (25%RTP (outside the applicable condition) within 4 hours.The current action allows 36 hours to place the plant in a non-applicable condition. As such, this is an additional restriction on plant operation. TECHNICAL CHANGES-LESS RESTRICTIVE"Specific" Ll The requirement to initiate action within 15 minutes to restore the limit is relaxed and relocated to the Bases in the form of a discussion that"prompt action" should be taken to restore the parameter to within limits.Immediate action may not always be the conservative method to assure safety.The 2 hour completion time for restoration of the limit allows appropriate actions to be evaluated by the operator and completed in a timely manner.BFN-UNITS 1, 2, 8L 3 evision 0 PK'a'-W} UNIT 1 CURRENT TECHNICAL SPECIFICATION .MARKUP

~p f 4 z.z.FES 24 pe 3.5.R 4.5.X 3"ski 4lcekJow 4r C4 pe PFN is~e.z.2 4.5.K.2 (Cont'd)b.4 as defined in the CORE OPEI4LTIHG LIMITS REPORT folloving the conclusion of each scram-time surveillance test required by Speci-fications 4.3.C.l and 4.3 AC.2.The determination of the limit must be completed vithin 72 hours of each scram-time surveillance required by S ecification 4.3.C.PppIi~4;q Lco S.z.g jLcrggd A Whenever the core thermal pover is g 25Z of rated, the ratio of FRP/CMFLPD shall be g 1.0, or the APRM scram setpoint equation listed in Section 2.l.h and the kPRM rod block setpoint equation listed in the CORE OPERATISQ LIMITS REPORT shall e multiplied by FRP/CMtLPD. When it is determined that 3.5.L.l is not being met, 6 hours is alloved to correct the condition. se~2 PRP/CNPLPD shall be determined daily vhen the reactor is g 25Z of rated thermal pover.Add profo Sf Pig g~7 4jz 3.g.g 1 H~P~~SR W.z.C~AcnaN 8 3o If 3.5.L 1 and 3.5.L.2 cannot be met, the reactor pover shall be reduced to 25Z of rated thermal pover thin 4 hours.BPS Unit 1 3.5/4.5-20 lilENVmr N,.2 6 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

FEB 2 c>sss 4, Appi'c~L.I, fj.LCc 3Zq whenever thc core thermal'pover ia g 25%of rated, the ratio of FRP/CMFLPD shall be g 1.0, or the APRM scram setpoint equation listed ia Section 2.l.k and.the kPRM rod block setpoint equation listed ln the CORE OPERATIHC LIMITS REPORT shall be multiplied by FRP/CMFLPD. SR L FRPICNPLPD shall bs determined daily when the reactor ia g 25K of rated thermal povcr.Ad<pro~~lsP F~~~y ag xR 3.2,A.(HZ.P~pe~S'a F24 z.P cTigg A 2~When it is determined that 3.5.L.I is not being mct, 6 hours is alloved to correct the condition. 30 If 3.5.L.l an&3.5.L.2 cannot bc met, the reactor pover ahall be reduced to 25X of rated thermal pover thia 4 hours.~p.s 1.The reactor shall aot be operated at a thermal pover and core flov inside of Regions I and II of Figure 3.5.M-1.2 If Region I of Figure 3.5.M-1 ia entered, immediately iaitiate a manual scram.3 If Region II of Figure 3.5.M-1 is catered: 1.Verify that.the reactor is outside of Region I and II of Figure 3.5.M-1: a.Follovlng any increase of more than 5X rated thermal povcr while initial core flov is less than 45X of rated, and b.Follovlng any decrease of more than 10K rated core flov while initial thermal pover ia greater than 40K of rated.BPK Ualt 2 3.5/4.5 20 NEWNESS NN, 282 WGE Ol UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP FEB 2 4>m'Co 3.2.Q&<AN A (l.)Whenever the core thermal (pover is g 25Z of rated, the ratio of FRP/CMFLPD shall be?1.0, or the hPRM scram setpoint equation listed in Section 2.1.k and the APRM rod block setpoint equation listed in the CORE OPERkTIHQ LIMITS REPORT shall be multiplied by HP/CNtLPD. When it is determined that 3.5.L.1 is not being, met, 6 hours is alloved to correct the condition. Sg.hi 4/CMFLPD shall be determined daily vhen the reactor is g 25Z of rated thermal pover.P~s.Pre~~~/4~df-Sg gap)H>P~p~5Q 3~If 3'.L.1 and 3'.L.2 cannot be met, the reactor pover shall be reduced to 25Z of rated thermal pover thin 4 hours.BPK Unit 3 3,5/4.5-20 AMENWar eo.r 90 PAGE~OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.2.4-APRM GAIN AND SETPOINTS ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specification, NUREG 1433.These changes should make the BFN Technical Specifications easier for the operator (and other users)to read and understand. During the reformatting and renumbering process, no technical changes (either actual or interpretational) were made unless they were identified and justified. A2 The current LCO and the proposed ISTS LCO ensure acceptable operating margins by limiting excess power peaking or reducing the APRM flow biased neutron flux upscale scram setpoints by the ratio of the fraction of rated power and the core limiting value of the MFLPD.Proposed ISTS LCO Item c also provides the option of increasing the APRM gains to cause the APRM to read a 100 times MFLPD (in%).This condition is to account for the reduction in margin to the fuel cladding integrity safety limit and the fuel cladding 1%plastic strain limit.Either a gain adjustment on the APRMs or an adjustment to the APRM setpoints has effectively the same result.Although BFN Technical Specifications do not specifically call out APRM gain adjustments, they are interpreted as an acceptable alternative and are allowed by current BFN plant procedures. For compliance with proposed LCO Item b (APRM setpoint adjustment) or Item c (APRM gain adjustment), only APRMs required to be OPERABLE per proposed LCO 3.3.1.1 (RPS Instrumentation) are required to be adjusted.A3 The CTS requirement (CTS 3.5.L.3)to reduce power to<25%of rated thermal power within 4 hours has been changed<25%of rated thermal power consistent with the LCO applicability for the CTS and the proposed BFN ISTS.The intent of the CTS action is to exit the LCO applicability and obviously this cannot be done until power is reduced below 25%.The change is slightly more restrictive by the literal wording of the technical specifications, however, since it does not represent an actual change to the intent it has been classified as an administrative change.TECHNICAL CHANGE-MORE RESTRICTIVE A new Frequency has been added to require verification of MFLPD within 12 hours of reaching or exceeding 25%RTP.This is an additional restriction on plant operation. CTS would allow up to 24 hours after reaching 25%RTP to perform the test.BFN-UNITS 1, 2, 5.3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.2.4-APRM GAIN AND SETPOINTS H2 A new Surveillance Requirement (SR 3.2.4.2)has been added that specifically requires the licensee to verify that APRH setpoint or gains are adjusted for the calculated HFLPD when the method of complying with the LCO is to make these adjustments. Since this change adds a specific requirement where none existed before, the change is considered more restrictive. BFN-UNITS I, 2, L 3 Revision 0 0-l0 JUSTIFICATION FOR CHANGES BFN ISTS 3.2, POWER DISTRIBUTION LIMITS BASES, The Bases for the current Technical Specifications for this section (3.5.I, 3.5.J, 3.5.K, and 3.5.L)have been completely replaced by revised Bases that reflect the format and applicable content of proposed BFN Unit 2 Technical Specifications Section 3.2, consistent with NUREG-1433. The revised Bases are as shown in the proposed BFN Unit 2 Bases. Enclosure V Volume 1S 3Pec 0'&on 3.'f, l NOV 18 1S88 4.6.D.R V v 588 3~+A~capon 4<<kz<gcs*< 8P<l ST 5 Z.g.2,~ggqg 3.The integrity of the relief valve bellovs shall be continuously monitored when valves incorporating the bcllovs design are installed. 3 6 E 2aMmua 4.ht least one relief valve shall be disassembled and inspected each operating cycle.E~Rum'.Whenever the reactor is in the SThRTUP or RUN modes, all)et yumya shall be operable.If it is determined that a)et pump is inoperable, or if tvo or more get pump flov instrument failures occur and cannot be corrected vithin 12 hours, an orderly shutdown shall be initiated and the reactor shall be placed in thc COLD'SHUTDOWN COHDITIOK vithin 24 hours.1.Whenever there is recirculation flov vith the reactor in tha STARTUP or RUE aodaa vith both recirculation pumps running, 5et yap operability ahall ba checked daily by verifying that the folloving conditions do not occur al simultaneously: ~ased log, VC'fig r SRZV.(,l~SR.,I 4~va racir at>~loo taaa-e-flo~~~~or<z.Ius aoea when'-i%~opera" b.Ink~I~l~The indicated value of core flov rata varies from the value derived froa looy flov measurements by sore than 10K.BFl Unit 1 3.6/4.6-11 c.The.diffuser to lover plan~di f ferantial pressure reading on an individual)ct yump varies from the mean of all]ct yump differentia.'ressures by aors than 10K.AMBINENT IjL Q~ 'i AUG 0 4 5$~<<5'uS44'r'Cab'O~ P~gQ~S 4 BFd I STS Z,q,~Pr~d Qp~6$R g...)4.6.Z.Whenever there is recirculation flov vith the reactor in the SThRHJP or RVH Mode and one recirculation pump i is operating, the diffuser to lover plenum differential pressure shall be checked daily and the differential prcssure of an individual)et pump.in a loop shall not vary from the mean of all get pump differential pressures in that looy by more than 10X.i&03igel+m r 1.The reactor shall not be operated vith one recirculation looy out~r;oeie of eerrfoe for sore theo~boors Vith thc reactor operating, if oac recirculation loop is out of service, the ylant shall bc placed in a HOT SHUTDOWNS CONDITION vithia 24 hours unless~the loop is sooner returned to.scrvicc.s, II 2 Recirculation pump s eeds shall be chewed e at least ce pcr ay.2~Fol oviag c y th dis ge va sp ed p may css t e spec p is css atsd s aed opera on, e of e lov t be o ened of th faster 50K of its 3~rgC7'>od D Mhaa the rca is n in thc CUR aode, CTOR POWER OPERATION vith both rec rcu at on pumps out-of-service for uy to 12 hours is permitted. Dur ag such interval estart of the recirculation yumps is permitted, yrovided thc loop discharge temperature is vithia 7$'F of the saturation 3.Bcf rc star ing cit r re rculat on ump d ing CT0 PO R IO, cck og th loo dis e teapc tur and do~at ati t cra ure.BEE Unit 1 sec 3MHg'~h'o 4 Qe~gds Pr ggh)(5T5 g tf 3'/4.6-12 AMENDMENT NO.2 g 7 f k.oo QvioN temperature of e reactor vessel vatcr aa determined by d pressure.The tota elapsed time natural circulation and one pump operation must be no greater than 24 hours.4.The reactor shall not be operated vith both recirculation pumps out-of-service vhile the reactor ia in the RUB mode.Polloving a trip of both recirculation pumps vhile in the RUN mode, imnediatcly inftiate a manual reactor scram.3.6.C 4+6.C The structural integrity of hSNE Code Class 1, 2, and 3 equivalent components shall be maintained in accordance vith Specification 4.6.G throughout the life of the plant.a.Vith the structural integrity of any ESNE Code Class 1 equivalent component, vhich fa part of the primary ayatca, not conforming to the above requirements, restore the structural integrity of the affected component to vithin ita limit or maintain the reactor coolant system in either a Cold Shutdovn conditfon or less than 50 F above the ainfiime temperature requfred by HDT considerations, until each indication of a defect haa been investigated and evaluated. 1.Inservice inspection of ESNE Code Class 1, Class 2, and Class 3 components ahall be.performed in accordance vith Section XI of the hSNE Boiler'and Pressure Vessel Code and applicable Addenda aa required by 10 CFR 50, Section 50.55a(g)j except vhsre specific vritten relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.hdditional inspections shall be performed on certain circumferential pipe vclda to provide additional protection against pipe vhip, vhich could damage auxiliary and control systems.BFS Unit 1 3.6/4.6-13 AMEKOMHP go p 06 0 S'F'.inca. 'on 3.M HAY 3 1594<CO Z.Q.)l.the reactor shall not be operated at a thermal pover and core flov inside of Regions I and II of Figure 3.5.N-l.2.If Region I of Figurc 3.5.N-1 is entered, faaedf ately initiate a aanual scram.SR Verify that the reactor is outside of Region I and II of Figure 3.5.5-1I a.Folloving any increase of aors than 5Z rated , theraal pover vhile initfal core flov fs less than 45K of rated, and 4An 8 If Region II of Figure 3.5.8-1 is entered: a.Iaecdfately initiate action and exit thc re ion vithin 2 u ert rol ods or by in reasi c c flo (sta ting eci culat on p t t ere fon an a 0 fa e ac 00)LR~b.While exiting the region, iaaediately initiate a manual acre%if thcraal-,hydraulf c instability is obserred, as cridcnced by o illa iona ch ed 1 percen pe to-p of r ted or LPRN acfll tions ch exes 30-pc ent peak-t peak seal.If peri ic LP ups e.o Qqps e ala oc ,'atel che the APRH's and indi idual RN's r erid ce of.the-hydraulic tab ity.b.Folloving any decrease of aors than 10K rated core flov vhile initial thermal pover is greater than 40K of rated.Pcopsr d Qc+'o n BlK Unit 1 3.5/4.5-21 NENbggPN.2p g

r~~~ill IIIIII I.~o UNIT 2 CURRENT TECHNICAL SP ECIF ICATION MARKUP 0 4 S S 0 5 cci figqgjo~g g/HOV 18 1988-'NITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS+L 4~sfili(qfio ~4r No~g~gg~Bylaw ISIS g yZ 3.g 3 3~4~The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. At least one relief valv shall be disassembled and inspected each operating cycle.3.6.E.~Jt Pupas E.~Jet um Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is determined that a jet pump is inoperable, or if two or more jet pump flow instrument failures occur and cannot be corrected within 12 hours, an orderly shutdown shall be initiated and the reactor shall be shutdown in the COLD SHUTDOWN COHDITIOH within 24 hours.~~5~ti 4/.Whenever there is recirculation flow with the reactor in the STARTUP or RUH modes with both recirculation pumps running, jet pump operability shall be checked daily by verifying that the following conditions p~~~g do not occur*-sa~.V.l.i simultaneousl ~v'cr.C'Lb a.The%vo ec rcu ation loopg%ave-a flow~:s~gt;>or ii ge kgb~.when.~opera~he-The dicated value of core flow rate varies from the value derived from loop flow measurements by more than 10X.h c,e J.4,!OOP c.The diffuser to lower plenun differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than lOX.BFH Unit 2 3 6/4.6-11 hMMmetn.%la

A'l AUG 04594 3854AL40~fol+a Q phJ Ql.l Pape@Nk~sR p.g.f.i 4'.E.2~Whenever there is recirculation flov vith the reactor in the SThRTtJP or RUR Mode and one recirculation pump ia operating, the diffuser to lover plenum differential pressure shall bc checked daily and the differential pressure of an individual)et pump in a loop shall not vary from the mean of all)et pump differential pressures in that loop by more lOX.QAI LCO 8.4)a4.44 Qo~~~:~~RZ.1.The reactor shal not be operated vith one recirculation loo out of service for more 24 hours.With the reactor operating, one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWR COHDITIOR vithin 24 hours unless the loop ia sooner returned to service.~0~9.V.//Recirculation pump ayceda M2.shall be checked A>at least once per A3 Al Pollov ng one pump operation, the d achargc va e of the v ape yump may t be open sa thc sp d of the f ter y ia less SO%of its eated speed.30 QVIOQ.D,,'Whan tha reactor ia not in the RJJK mode GTOR POWE HDtATIOR o recircu-on yea out~f-service for uy to 12 hours ia crmitted.au interval, restart o thc racirculation pumps ia permitted, provided the loop diachare temperature ia vithin'5 2'f the saturation temperature of the reactor 3~R2 Bcfor starting, either Foci ation dur REi POWER 0 OE, ck 1 thc lo y dia gc tcmyerat c and d saturation temperature ~BtÃUnit 2 c 5th W~d'IfiCagi~ ~~<C4~gg 4~BF<IsM z.9 9 3~6/4.6-12

Z eE<<$,.>.V.I NR i 8 1993 3~~gcgm8 h vessel water as determined by dome pressure The total e apsed time in natural circulation and one pump+I operation must be no greater than 24 hours.Secs.~);C;,.4;og. CE 4a<SftJ Lszg 4.pgsiod E The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.Following a trip of both recirculation pumps while in the RUN mode, ismediately initiate a manual reactor scram..G'.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to thc above requirements, restore the structural integrity of thc affcctcd component to within its limit or maintain thc reactor coolant system in either a COLD SHUTDOWN CONDITION or less than 50'F above thc minimum temperature required by NDT consider-ations, until each indication of a defect hae been inves<<tigatcd and evaluated. Inservice inspect,ion of ASIDE Code Class 1, Class 2, and Class 3 components shall be performed in accordance witt Section XI of the ASME Boil~and Pressure Veseecl Code anc applicable Addenda as requi: by 10 CFR 50, Section 50.55c except where specific writtc relief has been granted by 1 pursuant to 10 CFR.50, Sect: 50.55a(g)(6)(i). 2.Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.BFÃUnit 2 Se~~iS4eyio-4r C~~-gez*~cw5 Z.4.p//Q g i)~"'~Sec~ii<3 6/4~6-13 AMENDMENT lE 8 0 6 PAGE

Ai FEB 2 4 1995 L.t o L.l.Whenever the core thermal pover is g 25Z of rated, thc'ratio of FRP/CMFLPD shall bc Z 1.0, or thc APRM scram sctpoint equation listed in Section 2.1.k and the APRM rod block setpoint equation listed in the CORE OPEKLTIHG LIMITS REPORT shall be multiplied by FRP/CMFLPD. FRP/CMFLPD shall be dctermincd daily vhen the reactor is g 25Z of rated thermal pover.Qg J~g k Cite,ly~fi r CA'<7Q 4~8~N isis z.z.f 2.Shen it is determined that 3.5.L.1 is not being, mct, 6 hours is allovcd to correct the condition. 3.If 3.5.L.1 and 3.5.L.2 cannot bc mct, the reactor povcr shall be reduced to g 25X of rated thermal pover vithin 4 hours.Ai/CO 1.Thc reactor shall not bc operated at a thermal povcr and core flov inside of Regions I and II of Figure 3.5.N-1.2.If Region I of Figure 3.5.N-1 A is entcrcd, immediately initiate a manual scram.Sg'.0.I-2.1.Verify that the reactor is outside of Region I and II of Figure 3.5.N-1: a.Follovtng any increase of more than SX rated thermal povcr vhile'nitial core flow is less than 45X of rated, and 3.If Ze~g II of.Figurc 3.5.N-1 is'entered: b.Follovtng any decrease of more than lOX rated core flow vhilc initial thermal povcr is greater than 40X of rated.BFH Unit 2 3'/4.5-20 aemoMapgo. 2S2 5',- N OO 3~~~a.yq IG~8 Immediately initiate action and exit the region vithin 2 hours nsert ng con ro o s or b increas ng cor f v.('tar g a re rcu-1st~n pump exit t region s~ot an appropriate action)and b.While exiting the region, immediately initiate a manual scram if thermal-hydraulic instabilit is obse as evidenc d by AP oscil a-ti ns vh ch excee 10 p cent pea-to-p ak of r ed or PRN osci atio vhich exceed 30 pe ent eak-to-eak of cale.'Cf p riodic LP3X scale r d vnscale alarms oc r, edi ely ch ck the APRM and ndi idual ARM's for e denc of thermal-hydraulic ins ability.BFN Unit 2 3;5/4.5-20a

100.3.V.(-(Fi ure BEN Power I=low Stabilii:y Regions Q~00000 9~~0 000~0~0~0~~0~0~~100%Rod Line.R 0 CJ O I 0 C 6)O 0)CL ID 0 CL 0)0 (3 80-70-~~0~~0~~0~000 f 000~000~g~0~0~0~0~~\~0~~0~~0~50.~~~~40-.30--.Noturol (:irculotion, Line Q 00 0~~~00~~~~2 10-~~~~oo~0~~~~~~~~00~0~~0 0 PP~I 5 10 01~tW 15 20 25 30 Note: OperoVion Not 60;----Permitled in TI>is Region 807.Rod Line~~~eend~~~~~~\~~~~~~~~~~~~0~0~~~~~~0 l[jggm Raijion Rnoion Io~~~~~~~~~~~~~~r~~rm r~<~~35 40 45 50 55 60 65/0/5 80 85 90 95 100 10 Core Flow (perrcnt of rnted)CD C3 CTl 5 0' UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP NG CO TONS FO 0 ON SPeC Plea.OY 18 tS88 SUR C See 3WS+06CC fjOn 4e~eS+r tlPA 1sT5 s.e.x~30 9e 3.The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. 4.At least one relief valve shall be disassembled and inspected each operating cycle.3.6.E.J~ee Pum E.J~e~im~1.Whenever the reactor is in the STARTUP or RUN modes, all jet pumps shall be OPERABLE.If it is determined that a jet pum'p is INOPERABLE, or if two or more jet pump flow instrument failures occur and cannot be corrected within 12 hours, an orderly shutdown shall be initiated and the reactor shall!be placed in the COLD SHUTDOWN CONDITION within 24 hours.583.'].I-I Whenever there is recirculation flow with the reactor in the STARTUP or RUN modes with both recirculation pumps running, jet pump operability shall be checked daily by verifying that the following conditions do not occur simultaneously: +~'y a X~vo ecirculati oops hes~a flow~(s]of or 4504K.when l]4', 1 6C~opera tC'E.E nu+lc El b.The indicated value of core flow rate varies from the value derived from loop flow measure-ments by more than lOX.0 BFN Unit 3 3.6/4.6-11 c.The diffuser to lower plenum differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10K.AIAENDINBII'5. 1 9 9~~a=..& 0 h BO ARY Ai QPgc'gjcctM'1 3b/~(AtJS 04594 5Ll564i4AW oA 6<Cl~es e Bknl t S TS Z.9-'2-L f TO)05ed@ok 6 5R 3A.I.I 4 6 E.~Jam I 2.Whene ver there is recirculation flow with the reactor in the STARTUP or RUH Mode and one recirculation pump is operating, the diffuser to lower plen~differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all get pump differential pressures in that loop by more than 10K.i co z.e.(R4c eeerrrrm rrrr~lrr2 1.The reactor shall not be operated with one recirculation loop out of service for more than 4 hours.With the reactor operating, if C+g one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH within 24 hours unless the loop is sooner returned to service.SR3.~I l 1.Recirculation pump speeds shall be checked'm 1 at east nce er a Laz.2.~LA I F liow ng e-p p o era ion, e sch ge alve of e 1 pe p m no be pene un ss e eed f e f te p p i le s t 5 of ts ate spe d.3.ACTIN al D When th eactor is not-'in th RUH mode, REACTOR POWER OPERATIO with both recirculation pumps out-of-service for u to 12 hours is permitted. uring such interval estart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature 3.Bef e st rti eit er re rcu tion pum d ing CT P R ERA OH, e and og e lo p d cha ge tern ratu e do e saturation tempe ture.BFN Unit 3 See yooiiCrcrHon for<largP 6" BFN~SYS Z.H.9 0 Agio g D of the reactor vessel water determined b dome ress The tota e apsed time fn natural circulation and one pump operation must be no greater than 24 hours.See 5~sWWog fc~~~<krBPN igTsp q.q, 4~AC4'oui, E The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G S ctu 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1"equivalent component, which is part of the primary system, not conforming to the above r'equirements, restore the structural fntegrity of the affected component to within fts limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50'F above the miniinm temperature required by HDT consfder-ations, until each indication of a defect has been investigated and evaluated. l.Inservice inspection of ASME Code Class 1, Class 2, and Class 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, Section 50.55a(g except where specific written relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additional inspections shall b performed on certain circumferential pipe welds to provide additional protection against pipe whip which could damage auxiliary and control systems.BFH Unit 3 See S~W'cOHo+6~ch~ys 6 cps s.<%6.r..G in yh;S geC4og 3.6/4.6-13 ANBMgr NO.y 79 paG 0 ~~~~;~:w,~ats&e1iJBYilt(hsrtit&eloeily7%(lMJt Ram~qn w;I~P$:~.'i+igl J g4'A>4f yy~~XI)'EMIIHIIT') CtIi1'J Sa'e->~~~~'~~~~~~~~~~~~~~~~~~~~~'Q'~~~~~~I~I~~~~I~~~~~~~o~.~g~~,~I~I~I~~~~~~~~I~~~'I~~~'~~~I~~4t~~.I~~~~~~~~~~~I~~~~~~~~I~~~~~I a~~I~~~>~~I~~~1~~~~~~~~~~~'I>~~~~i~~

~~~I~~~~~ E JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in'a technical change.CTS requires the plant to be placed in the HOT SHUTDOWN CONDITION in 24 hours with one recirculation loop out of service.Proposed ACTION C requires the loop be returned to service in 12 hours or ACTION D requires the plant to be in MODE 3 (Hot Shutdown)in 12 hours.The CTS and the proposed ISTS Completion Times are essentially equivalent since both require the plant to be in MODE 3 in 24 hours.A3 The frequency for this Surveillance has been changed from once per day to once per 24 hours.This is a terminology change and is therefore administrative. TECHNICAL CHANGE-MORE RESTRICTIVE CTS allows up to 24 hours operation with the reactor power<1%with no recirculation loops operating (the total elapsed time in natural circulation and one pump operation must be no greater than 24 hours).Proposed ACTION D is more restrictive since the time limit of 12 hours applies to<1%while in MODE 2 also.0 BFN-UNITS 1, 2, 5.3 Revision 0 Pi,GE Vt 3 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING H2 The flow imbalance limit is being reduced to 10%of rated core flow when operating at<70%of rated core flow, and to 5%of rated core flow when operating at a 70%of rated core flow.The current requirement is 15%mismatch of flow at the given flow conditions. While the limit appears to be less restrictive if core flow is x 66%of rated core flow, it is more restrictive when'>66%of rated core flow (i.e., 15%x 66%or less is x 10%of rated core flow), where the unit normally operates.In addition, currently, this is only a problem if there is an imbalance in combination with two other conditions (CTS 4.6.8.l.b and c).The new requirement is separate from the other two, thus, actions will now be required if there is an imbalance by itself.Therefore, this change is considered more restrictive on plant operations. TECHNICAL CHANGE-LESS RESTRICTIVE"Generic"~LA1 This requirement is being relocated to plant specific procedures. The purpose of this limitation is" to provide assurance that when shifting from one to two loop operations, excessive vibration of the jet pump risers will not occur.Short term excessive vibration should not result in immediate inoperability of a jet pump, but could reduce the lifetime of the jet pump.This type of requirement is generally found in plant operating procedures, similar to other operating requirements necessary to minimize the potential of damage to components. Changes to the procedures will be controlled by the licensee controlled programs.LA2 This requirement is being relocated to plant specific procedures. Details of the methods for performing this Surveillance, and any requirement to record data, has been relocated to plant procedures. Any changes to the procedures will be controlled by the licensee controlled programs.LA3 These requirements are being relocated to plant specific procedures. The details of the acceptable method for meeting an action requirement and what constitutes evidence of thermal hydraulic instability and the need to check for it have been relocated to plant procedures. Any changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 5 3 PAGE~OP Revision 0 Cl JUSTIFICATION FOR CHANGES BFN ISTS 3.4.1-RECIRCULATION LOOPS OPERATING"Specific" Ll This change adds a note which states the Surveillance is not required to be performed until 24 hours after both recirculation loops are in operation. The Surveillance is not required to be performed until both loops are in operation since the mismatch limits are meaningless during single loop'or natural circulation operation. Also, the Surveillance is allowed to be delayed 24 hours after both recirculation loops are in operation. This allows time to establish appropriate conditions for the test to be performed. L2 Per CTS 3.5.M.3.a, if Region II of Figure 3.5.M-1 is not exited within 2 hours, the Specification is violated and CTS 1.O.C.1 applies requiring the plant be placed in Hot Standby within 6 hours and in Cold Shutdown within the following 30 hours.This provides actions for circumstances not directly provided for in the specifications and where occurrence would violate the intent of the specification. The BFN ISTS provides Action within the Specification which could be considered less restrictive than CTS.Action 0 allows 12 hours to be in MODE 3 (Hot Shutdown)and 36 hours to be in MODE 4 (Cold Shutdown). The proposed Action is considered less restrictive since 12 hours is allowed to place the unit in Hot Shutdown versus the 6 hours allowed to place the unit in Hot Standby per CTS.BFN-UNITS 1, 2,&3 Revision 0 UNIT 1 CURRENT TECHNICAL SP ECIF ICATION MARKUP NOV Z8 1988 4.6.D.R e Va ve 5'C'('~c+',4;(c,h~ p+~$<o+<BP+ILATS 3,q, 3~The integrity of the relief valve bcllovs shall be continuously monitored vhen valves incorporating the belloys design arc installed. LCo 3.9.2 Whenever thc reactor is in thc Rgplicab'1'kj STARTUP or RUH modes, all,)et umps shall be operable.If it is determined that a jet pump is inoperablc, r tv or re t pump ov t cnt fa urea occur d c ot c rec d 12 ours an order y shutdovn shall bc initiated and the reactor shall'be placed in the~HUTDOWH COHDITIOH vithin ours.4~At least onc relief valve shall be disassembled and inspected each o erati c cle.Vcn c&oLht.o~(. of-lt~~i~P C~ehn~Mti Fi'CA (eLcg o~ee 0 A iowan enevcr there is ecircu ation flov vith he re ctor th TAR or cs ith 0th ecir Lt um s ingg ct op rabil s 11 b checked aily vcrifyi t th follov ng c diti do not occ simultaneously: ~e Y~ShA(cfhon 4,f-~~>gci gPN LS75 p,q.~frogoCd<R>'t>I IJokS g2 ftopsH SP Z.9.3.I a.The tvo recirculation loops have a flov imbalance of 15X or morc vhen the pumps arc op~rated at th s ccd b.The indic ted value of orc ov ate v ies rom e luc eriv f oop ov mess em ts more han 10K eke cc~]isa&%Pe v l.$.The dfffeeer eo lower plenum dif ferential pressure reading on an individual)et p varies from BFH Unit 1 3.6/4.6-11 C g less g2.NENOMENT NL I 5 8

Se>>iP;on 5.,z AUB 04 1994 3.6.F~>m Z~Von 4-C/g~~k~8pnl Isis z.g,~~se~lICab;fig 4.6.F ev r t ereyis cu ti A,o vith thc reactor in the~STARTUP or RUR Mode and n rec culat n um 8 diffuser to lover plenum differential pressure shall be checked a and thc differential pressure of an individual get pump.in a loop shall vary from m of all get p dif er tial (mesa rea t t o by eave~z~ass t}x 1.The reactor shall not bc operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if onc recirculation loop is out of service, thc plant shall bc placed in a HOT SHUTDOWN CORDITIOR vithin 24 hours unless the loop is sooner returned to service.1.Recirculation pump speeds shall be checked a'nd logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed pump may not bc opened unlesa the speed of thc faster pump is less than 50Z of its.rated speed.3.When the reactor%a not in thc RUE mode, REACTOR POWER OPERATIOR vith both recirculation pumps out-of-service for up to 12 hours is permitted. During such interval restart of the recirculation pumps ia permitted, provided thc loop discharge temperature is vithin 75'P of the satu 2.Ro additional surveillance required.3.Before starting either recirculation pump during REACTOR POWER OPERATIOK, check and log thc loop discharge temperature and dome saturation temperature. BFK Unit 1 3.6/4.6-12 AMENDMENT No.2 g y-:.3-

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~GP~ 0 ION s NOV 18 1S88 4.6.D.e Va ves 5'VC7 lgjCA7 IohJ Fog~ho/~/or B j=w 3~The integrity of the relief valve bellovs shall be continuously monitored vhen valves incorporating the bellows design are installed. Al~~~Lu)3.VZ I-~Apl'c4'.j,g Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is determined that a Jet pump is inoperable or i~two or ore get pump flov instibment failu s occur~cannot orrecte vithin 12 hours an orderly shutdown shall be initiated and the reactor shall be shutdown in the 66BB-HUTDOWH COHDITIOH vithi hou/2.ad QuST(@CA Tkr&FbA C4<u~84, BFH'ggg g g.]4~At least one relief valve shall be disassembled and inspected each'erati c cle.c Q ak o<c.C1.4'ra lS Sa'l<gg~CI h'4 i~g/enever t vk~/os recircul ion flov vith the re tor in the TAR or RUH modes it both rec culation p ps runni, get pump operabili shall be checked aily b veriiyi tha the follovfng c ditions do not occur a.The tvo recirculation loops have a flov imbalance of 1SX or more vhen the pumps are operated at the same s eed.4~~~SW Sa S.MZ.I No4S~b.The i icated value of c re flov ate var es from e.v ue deri ed ir m oop flo measurements by more BFH Unit 2 Prolog~Sg.3.Q 2./~dc SR 3.42.y$~~as'I~4/p ffcr~3.6/4.6-11 The diffuser to lover plena differential pressure reading on an individual et pump varie from by p./'MENDMEÃf IL I 54 Al 3.6.F SeeadÃc4m 4~a fc~BFN Is~~'I.J t 0 t D~sg 3.Cz,l Wh never there is reci c atioWflo vith]the reactor in the"'"'~M (STARXUP or RUH mode~one~ec~u at o~uidp o erati , t c diffuser to lover plenum differential pressure shall be checked pS@~ail and the differential pressure of an individual jet pump in a loop shall vary from e me o all)et ump-di ential essurc loop y than~C5 2d L3 4.6.F.1.The reactor shall not be operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if one recirculation loop is out of service, thc plant shall be placed in a HOT SHUTDOWH COHDITIOH vithin 24 hours unless the loop is sooner returned to service+1.Recirculation pump speeds shall be checked and logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed pump may not bc opened unless thc spccd of the faster pump is less than 50K of its rated speed.2.Ho additional surveillance required.3.When thc reacHmis not-in the RUE mode, REACTOR POWER OPERATIOH vith both recircu-lation pumps out-of-service for up to 12 hours is permitted. During such interval, restart of the recirculation pumps is permitted, provided the loop discharge tempcraturc.is vithin 75 F of thc saturation temperature of the reactor BFH Unit 2 3.6/4.6-12 3.Before starting cithcr recirculation pump during REACTOR POWER OPERATIOH, check and log the loop discharge temperature and dome saturation tcmpcraturc. AMENOMENr N.2 2 g.3..-3-- 0, UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 0! 5 e'A'cn',g.g HOV18 1S88 4.6.D.Re e Va ves See~~ggg~+~Shd IS TS yq.g 3~The integrity of the relief valve bellows shall be continuously monitored when valves incorporating the bellows design are installed. Ecole,g 2.enever the reactor is in the TARTUP or RUH modes, all jet~pumps shall be OPERABLE.If t is determined that a jet pump is INOPERABLE or or et pf wist nt fai ure occu and anno be orr cte with s an order y s utdown shall be initiated and the reactor shall be placed in the HUTDOWH CONDITION within ou Iz Ito7.~de+reeH~ifrrfroe gr Cluepr+~8PN.Is<5 g,q.i QA>4.At least one relief valve shall be disassembled and inspected each operating cycle.gtt 4 lcf36+o~6))auo'f~enev r recir ulati n fl w t the eact in he ST TUP R mo es w th bo re rc at n umps i , et p oper ilit s 11 che ed ily by veifyi t t t f liow ndi o o no oc r simultaneousl '.The two recirculatio loops have a flow imbalance of 15K or more when the pumps are operated at the same speed.J-Z Ifogsbcd 5R 3.9.g,(Plpkg b.e n at d lu of or fl w at v ie fr m he lu d iv d f om oo f w eas re-m ts y or than 1 X.sa e.~,>>The diffeeer re lever plenum differential pressure reading on an individual jet pump varies fire++ho flC csW0Q4e fh Hrea BFH Unit 3 3.6/4.6-11 than~40'/o la>>AMENOMHfT Ni)f 29,;P/s+t-~ ~~Ccs c~~AUS 0 4 594 5<+YuSkjkicaHon h r t:kpeg+~8~l SVS 8 N.1~~~SR cn er her is rc at on ov vith Ay~;cog],'Q~) the reactor in he C STARTUP or RUH Mo and e c ula i o c ati the di fuser to lover plenty differential pressure shall be che ail and the differential prcssure of an individual jet pump in a loop shall/}C va fro t m of a et ump h3 if cr tial s e t t o by move than 3.6.F e c at 0 a o 4.6.F t o 1.Thc reactor shall not be operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH within 24 hours unless thc loop is sooner returned to service.1.Recirculation pump speeds shall be checked and logged at least once per day.2.Folloving one-pump operation, the discharge valve of thc lov speed pump may not be opened unless thc speed of the faster pump is less than 50X of its rated speed.2.Ho additional surveillance required.3.When the reactor is not"in the RUH mode, REACTOR POWER OPERATIOH vith both recirculation pumps out-of-service for up to 12 hours is permitted. During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is vithin 75'F of the saturation temperature 3.Before starting either recirculation pump during REACTOR POWER OPERATIOH, check and log thc loop discharge temperature and dome saturation temperature. BFH Unit 3 3.6/4.6-12 a~arn~mt vo.i 8C JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUNPS ADN I NI STRATI VE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readil'y readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change..The wording of the surveillance was changed to require verification that one of the following criteria are met rather than verifying that none of the conditions exist simultaneously. This is consistent with NUREG-1433 which attempts to phrase everything in a positive manner.Due to the change in phrasing of the Surveillance,"more than" was changed to"less than or equal to" in criteria b and c.A3 The variance of the diffuser-to-lower plenum differential pressure reading on an individual jet pump will now be taken from the established pattern rather than from the mean of all jet pump differential pressures. This change is in accordance with the recommendations of SIL-330 and NUREG/CR-3052 and is consistent with NUREG-1433. A4 The conditions of the Surveillance Requirement are assured by LCO 3.4.1.Therefore, there is no need to restate the conditions for jet pump operability. A5 The frequency for this Surveillance has been changed from daily to once per 24 hours.This is a terminology change and is therefore administrative'. BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUMPS TECHNICAL CHANGE-MORE RESTRICTIVE Ml The requirement to place the plant in a Cold Shutdown condition within 24 hours when a jet pump is inoperable has been revised to reflect placing the plant in a non-applicable condition. Current Specification 1.0.C.1 states action requirements are applicable during the operational conditions of each specification. Therefore, the requirement to place the plant in Cold Shutdown is not applicable after Mode 3 is reached.The'revised action requires plant power to be brought to Mode 3 (outside the applicable condition) within 12 hours.The current action allows 24 hours to place'he plant in a non-applicable condition. As such, this is an additional restriction on plant operation which constitutes a more restrictive change.This change adds two requirements to the Surveillance to detect significant degradation in jet pump performance that precedes jet pump failure.The first requirement added would detect a change in the relationship between pump speed, and pump flow and loop flow (difference >5%).A change in the relationship indicates a plug flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle.The second requirement added monitors the jet pump flow versus established patterns.Any deviations >10%from normal are considered indicative of potential problem in the recirculation drive flow or jet pump system.These two added requirements to the Surveillance help to detect significant degradation in jet pump performance that precedes jet pump failure.Requirements added to Surveillance Requirements constitute a more restrictive change.In addition, CTS 4.6.E.1 allows jet pump operability to be verified by demonstrating that the two recirculation loops.have a flow imbalance of s 15%when the pumps are operated at the same speed.This is now a separate requirement (Proposed SR 3.4.1.1 See M2 of the Justification for Changes for Specification 3.4.1)and can no longer be used by itself to demonstrate jet pump operability. This change is consistent with NUREG-1433. 0 SIL-330 provides two alternate testing criteria (thus the deletion of current Surveillance 4.6.E.l.b).One method uses easy to perform surveillances with strict limits to initially screen jet pump operability (the proposed changes above).If these limits are not met, another set of Surveillances exist (current Technical Specifications). Revising the Surveillances to separate the flow imbalance test requirement and to include the stricter limits reflects a more restrictive change.BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.2-JET PUMPS TECHNICAL CHANGE-LESS RESTRICTIVE"Specific" Ll This change deletes the current shutdown requirement associated with jet pump flow indication. Currently, when required jet pump flow indication is lost, an orderly shutdown must be initiated in 12 hours and the reactor is required to be in Cold Shutdown within the following 24 hours (since Mode 3 is the non-applicable mode, then 24 hours is allowed to reach Mode 3;see discussion of change Hl for ITS 3.4.2).The proposed Specification implicitly requires the jet pump flow indication to be operable only for the performance of the Surveillance Requirement. If the flow indication is inoperable when the surveillance is required to be performed and jet pump flow can not be determined by other means, the jet pump would be decl'ared inoperable and the appropriate actions would be followed.Since the proposed jet pump surveillance requirement is required to be performed every 24 hours (the 25%extension per SR 3.0.2 can be applied)and the Required Actions require the reactor to be in Mode 3 within 12 hours, the maximum difference in the current Specification and the proposed specification is 6 hours.As a result, the proposed specification effectively allows a maximum of an additional 6 hours (which is the 25%extension) to reach a non-applicable Mode if a required core flow indicator is inoperable and jet pump flow can not be determined. Depending on when the failure occurs, 6 hours is the maximum increase over the current Specifications (failure occurring immediately after the surveillance is performed). The following table provides the details of the calculation of the 6 hour period: Current Tech Specs Time 0 hours-Jet Pump.Indication Fails-12 hr AOT Begins Time 12 hours-12 hr AOT Expires-24 hr AOT Begins to MODE 3 (per 3.0.A;see Ml)Time 36 hours-24 hr AOT Expires Plant in MODE 3 Proposed Tech Specs Time 0 hours-Jet Pump Indication Fails (Immedi ately After SR Time 30 hours-SR due;Flow (24 hrs x Indication Inop 1.25)-12 hr AOT to MODE 3 Begins Time 42 hours-12 hour AOT Expires Plant in MODE 3 BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR.CHANGES BFN ISTS 3.4.2-JET PUMPS L2 As depicted above, 42 hours is the maximum time that would be allowed if a required jet pump flow indicator is inoperable and jet pump flow can not be determined. Currently a maximum of 36 hours is allowed if more than one jet pump flow indicator is inoperable. Jet pump flow indication operability does not directly impact jet pump operability. Jet pump flow indication is only required to perform the jet pump Surveillance (SR 3..4.2.1).SR 3.4.2.1 verifies jet pump operability and has a frequency of every 24 hours.The 24 hours frequency plus the 25%extension has been shown by operating experience to.be timely for detecting jet pump degradation and is consistent with the surveillance frequency for recirculation loop operability verification. The most common outcome.of the performance of a surveillance is the successful demonstration that the acceptance criteria are satisfied. This change is consistent with NUREG-1433. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation. Note 2 to proposed SR 3.4.2.1 provides time to perform the required.Surveillance when the reactor exceeds 25%RTP.Below 25%RTP, low jet pump flow results in indication which precludes the collection of repeatable and meaningful data.The flexibility to proceed to a 25%RTP and then commence the SR every 24 hours is consistent with approved Technical Specifications for both Perry Nuclear Power Plant and River Bend Station.L3 The allowed difference between each jet pump diffuser-to-lower plenum differential pressure to the loop average has been increased to 20%.This change is consistent with the recommendations of SIL'-330 and NUREG/CR-3052 (Closeout of IE Bulletin 80-07: BWR Jet Pump Assembly Failure).SIL-330 specifies a 10/criteria for individual jet pump flow distribution. When measured by jet pump diffuser-to-lower plenum differential pressure, the equivalent limit is 20%because of the relationship between flow and delta-P.Since BFN uses the diffuser-to-lower plenum differential pressure measurement, the variance allowed should be 20%as recommended by SIL-330 and NUREG/CR-3052. This is a relaxation from existing requirements, therefore, it constitutes a less restrictive change.This increase in allowed difference is considered an acceptable criterion for verifying jet pump operability and is consistent with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 PAGE

UNIT 1 CURRENT TECHNICAL SPECIFICATION .MARKUP

OEC 0 7 1SS4 F 6.C 4'.C 2~Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours I for the air sampling system.,!2.With the air sampling sys tern inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.Ac 5'usaf'i~tjon gag Qqg HPFA)ST>pgq y~~~The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.3~Qai 1.A<Ho 8 R L2 If the condition in 1 or 2 above cannot be met, an orderly shutdoml shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.I 4Me3 in AH I>br s When o relief alve is kno~to be failed, an orderly shutdo~shall be i iti and the reactor epressurize o less than 105 w th hour e s ae ot eq red to b OP LE in e C IT N.SR 3.ge3ol A proxy e y one-half of all re ief va es sha 1 be b ch-ch ked or r laced with a eck ve A 2 each operatin c cle.All val es 11 have en heck or r lac d up the co let on of ever ecnd c es l,243 tfopScl SR s.q',g,~SR7'l 3~2 In accordance vith Specification MM each relief valve shall be manually opened t'1 t e oco ples nd ac stic moni rs d str am of the alve i dica e ste is lookin f the a lve.BFN Unit 1 3.6/4.6-10 AMENDMENT NL 2 Z 3 0 gkl NOY 18 1988 3.e i tegr y of the elie val e bel ovs shal be ontin ousl mo tore vhen alv in orpo ating he b 1 ws desi r tall d.At leas o rel ef al e s all e sass bl d d i p ted ach oper ti c e.3 6 E Jm~muu E.~Jet 1.Whenever the reactor is in the STARTUP or RUH modes, all get pumps shall be operable.If it is determined that a)et pump is inoperable, or if tvo or more Jet pump flow instrument failures occur and cannot be corrected vithin 12 hours, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.1.Whenever there is recirculation flov vith the reactor in the STARTUP or RUH modes vith both recirculation pumps running, get pump operability'hall be checked daily by verifying that the folloving conditions do not occur simultaneously: gee rw+,'p;(qp<g~<+es tr>8cnr lSTS~s.z.r+P~ps a.The tvo recirculation loops have a flov imbalance of 15Z or more vhen the pumps are operated at the same speed.b.The indicated value of core flov rate varies from the value derived from loop flov measurements by more than 1OZ.BFH Unit 1 3.6/4.6-11 c.The diffuser to lover plenum differential pressure reading on an individual)et pump varies from the mean of all)et pump differenti pressures by more than 10Z.AMENOMENr N.Z g 8 p~GC a.' 0 SAPBTY I INIT 1.2 Reactor Coolant S stea Inte rlt f INITIN3 SAPETY SYSTEN SBTTIHQ 2.2 Reactor Coolant S tea Inte rlt Applies to llalts on reactor coolant systea pressure.Applies to trip settings of thc instruecnts and devices which are provided to prevent the reactor systea safety lialts fraa being exceeded.O~b8cl 1v8 To establish a ligilt below which the integrity of the reactor coolant systea is not threatened due to an overpressure condition. o~h ective To define the level of the process variables at which autccaatic protective action is initiated to prevent the pressure safety limit free" being exceeded.S ecificatlons A.The pressure at the lowest point of the reactor vessel shall not exceed 1,375 psig whenever irradiated fuel is in the reactor vessel.Cgz~gc 4 gC 4 I S TS, 3.o The limiting safety systea settings shall be as specified below: Liwiting Safety rotcctlve Action S tea Settin SR 2'f.3l h.Nuclear systea 1.105 psig+relief valves sg, sl open-nuclear (4 valves)systca pressure 1,115 psig+33.5 osl (4 valves)~1.125 pslg+83.8-k+psl (5 valves)B.Scraa--nuclear <1,055 psig systen high pressure BFN unit l 1.2/2.2-1 S UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 0 OEt;0 V 1SS4 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout providing a temporary monitor.2.Pith the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.5+8~ST(FICA7/QPJ~Q CPA+G:E-svo BFN)'sl-s gc/g+3.0.5 p<glot4 p L,h goo&5 lg2Q)),ca),;l Q When ore than one relief valve is kaovn to be failed an orderly shutdovn shall b initiate e react r depress o lyas 10~ig vithin ours.e c ar ot req ed to bc in COLD SHUTDOWN 7rapos~Qp3 hble.ha sas4~~3.If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall be initiated and the reactor shall be placed in the COLD SHUTDOWN COlQITIOK vithin 24 hours.sp s..~l tely one-half o~l?,he al relief val s sh 11 be.ch-cheched r repla d vith a p,Z bench-chewed valv each operating cyc JQ.1 13 valves vill have h!Ien chere~or replace~on the compMtion of every 2.In accordance vith/,Z.5g.4'f.3.> Specification.O.tR each relief valve shall be aanually opened unti c c co tic m tora o trc of the alv dicate team i flo ng om thc valve.BFR 3 6/4.6-10 AMENDMENT 10.2 29 Fia=-~OF~ I HOY 18 1S88~r~~4 3 Tho r te'ty of che eh on be ont o ly to v es rpo atQxg the bellove design are install 4, leaa one r ief alv e di emb d azicl act ea the e.3.6A.&~max 1.Whenever the reactor is in the STOUP oz EHf modes, all Jet pampa ahall be OPELQKZ.it ia determined that a get pap is inoperable~ or if tvo or more)et pump flov instrument failures occur end cannot be corrected vithin 12 hours, sn orderly shntdovn shall be initiated and the reactor ahall be ekan~rn in the COLD SKFRtNN[-CanamOI vithin Zi~.QQ Qug4s ls qgL>0M C4c.~q*Q,Phd<gP~2 4 2, a e,k P-1 ps.1.Rxenever there is recirculation flov vith the reactor in the STQCUP or RON modes vith both recirculation poaye ramduS, get pmnp operability shall be checJcack daily by veri~that the f ollcndxw conditions do not occur simnltaneoasly1 a.The tvo recirculation loops have a flov imbalance of LSZ or more vhen the pampa are operated at the same speed+b.~indicated value of core flov rate varies from the value derived from loop flov measurements by more than 10K.c~The diffnaer to lover plenum differential pressure readfilg on an individuaJ, p1mep varies from the mean of all)et pump differential pressures by more than 10".Vn't" 3.h/4.6-11 AMENDMENT NO 15 4 SAFETY LIMIT l.2 Reactor Coolant S stem Inte rit LINITING SAFETY SYSTEN SETTING 2.2 Reactor Coolant S tern Inte rit Applies to limits on reactor coolant system prcssure'~e ective To establish a limit below which the integrity of the reactor coolant system is not threatened due to an overpressure condition. Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits frcci being exceeded.O~e votive To define the level of the process variables at which automatic protective action is initiated to prevent the pressure safety limit from being exceeded.. S ecifications S cifications A.The pressure at the lowest point of the reactor vessel shall not exceed 1,375 psig whenever irradiated fuel is in the reactor vessel.The limiting safety system settings shall be as specified below'ijiiting Safety protective Action S stem Settin 5R 9.0,9./h.Nuclear system 1.105 psig+relief valves 93 psi open-nuclear (1 valves)system pressure SEE'Q5TIF'Ic/TIoAJ F'ag CPANg gee~O gyral i~g Z~.115 psig+-3i.S (i valves)1,125 psig+si (5 valves)B.Scram-nuclear<1,055 psig system high pressure BPN Unit 2 1.2/2.2-1 PAGE UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP PAGE~OF~ 0 5 DEt'7 1994~6.C 4.6.C 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling'systems shall be OPERABLE'rom and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.2.With the air sampling system inoperable, grab samples shall be obtained and analyted at least once every 24 hours.~~X~564ie4on 0~chases+BIN jsTs 3'.q.y p3.V.s'he air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.od<$lp X43+l jCR bIll ProPOSC Jj HO<~SP.3.la.>3.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.'in M>jn jIhfS 1.When than on relief valve is known to be failed, an orderly shutdam shall b initiate d the reactor depressurite o lcaa 0 sx v thin~e LR e c s a n re i ed t e 0 ERAB i th CO CON TIO sg P.f>.sR z.s.s.c 2 proximately onc-half o 11 relief val s shal e bench-chec d or repl d vt,th a bench-ch eked vc, ach o r n c cle hll 13~alves vill have be checked r repla d upon t complet of eve second cycle.If'n accordance with Specification each relief valve shall be manual o ed t thermoco es and aco tic monito downs earn of the alve indicat steam is the valve BFH Unit 3 3.6/4.6-10 AMENVHrr NIL X 86 Po.GE~O'~ 0 SPecif)cafjnr) 3.Q.3'OV 18 t988 Q5 3.Th in egrity of the r ie val e b 11 vs~al bc ont nu usl on ore vh alv s inc rpo ati e ellovs de ign rc ns al d.4.At eas one rel ef v lve s ll di ass ble d i pcc ed ach 0 cra ing cyc~3.6.E.Jet~ups 1.Whenever the reactor is in the STARTUP or RUH modes, all jet pumps shall be OPERABLE.If it is dctermincd that a jet pump is IHOPERABLE, or if tvo or more jet pump flov instrument failures occur and cannot be corrected vithin 12 hours, an.orderly shutdown shall be.initiated and thc reactor shall be placed in the COLD SHUTDOWN COHDITIOH vithin 24 hours.E.J~e 1.Whenever there is recirculation flov vith the reactor in the STARTUP or RUH modes vith both recirculation pumps running, jet pump operability shall be checked daily by verifying that the folloving conditions do not occur simultaneously: 5'ee Tus&ceh>>n J>>~c4ngc's 4 BcN 7srs g.q.2 5<fpuw p g a.The tvo rccirculatio loops have a flov imbalance of 15X or more vhen the pumps arc operated at the same speed>>b.The indicated value of core flov rate varies'from the value derived from loop flov measurc-mcnts by more than 1OX.BFH Unit 3 3.6/4.6-11 c.The diffuser to lover plenum differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10X.AMENOMBfr N.I 29 5fec)@ca an 3'.<3 1.2 2.2 hpplles to liaits on reactor coolant systea prcssure.Applies to trip settings of thc instruments and devices vhich are provided to prevent thc reactor systca safety liaits froa being excccded.To establish a liait belov vhich thc integrity of thc reactor coolant systea ls not thrcatencd due to an ovcrpressure condition. To define thc level of thc process variables at vhich automatic protective action is initiated to prcvcnt thc pressure safety lialt froa being exceeded.h.The prcssure at thc lovcst point of the reactor vessel shall not exceed 1,375 psig vhenevcr irradiated fuel is ln thc reactor vcsselo The liaiting safety systea~ettings ahall bc as specified bclov!5R S4,.I J.Nuclear systea 1 105 pslg g relief valves 33,xM psi open-nuclear" (4 valves)systea pressure 115 craig g'f m S psl 4 valves)Scc SLL5+s gasw 4s f5'2eg 1 125 psig g.g~psi (5 valves)B.Scraa-nuclear g1,055 psig systea high prcssure BPK 1.2/2.2-1 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF VALVES ADMINISTRATIVE A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The Frequency for proposed SR 3.4.3.1 (CTS 4.6.0.1)has been changed from"each operating cycle" to"18 months." Since an operating cycle is 18 months these are equivalent. The Frequency for proposed SR 3.4.3.2 (CTS 4.6.0.2)has been changed from"In accordance with Specification 1.0 HH" to"18 months." Since the Inservice Testing Program (1.0.HH)frequency is 18 months these are equivalent. As such, these changes are considered administrative. A3 The proposed change adds a note that states that the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASHE code requirements, prior to valve installation. As such, the addition of the note is considered administrative. A4 CTS 3.6.D.1 requires an orderly shutdown when more than one relief valve is known to have failed.Therefore, the CTS allows unlimited operation with one S/RV inoperable. BFN has 13'/RVs, therefore, 12 are required OPERABLE at all times.LCO 3.4.3 requires 12 to be OPERABLE and shutdown if one of the 12 required S/RVs is inoperable. As such, the two Specifications are equivalent and this change in presentation is considered administrative. BFN-UNITS 1, 2, 5.3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF VALVES A5 BFN CTS 4.6.0.3 is only applicable to three stage Target Rock S/RVs.Only the two stage Target Rock S/RVs are'nstalled and authorized for use in BFN Unit 2.The three stage design is obsolete and is no longer supported at BFN.Since this Surveillance Requirement is no longer applicable to the BFN S/RV design, the deletion of this requirement is considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE Ml Ah'additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because't stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.6.D.1).CTS requires a shutdown to MODE 4 within 24 hours but does not stipulate how quickly MODE 3 must be reached.Reference Comment L2 which addresses the less restrictive change of be in MODE 4 in 36 hours rather than 24 hours.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl The details relating to methods of performing Surveillances have been relocated to the Bases or procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs.LA2 This Surveillance Requirement has been relocated to plant procedures since the requirement does not directly relate to S/RV operability. This is strictly a preventive maintenance requirement. 0 QF Pr."-BFN-UNITS I, 2, 5 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.3-SAFETY/RELIEF YALVES"Specific" Ll The allowed lift setpoint tolerance has been increased from 1%to 3%based on incorporation of this larger setpoint tolerance in the BFN reload licensing analysis for each Unit prior to ISTS implementation. The larger setpoint tolerance has already been incorporated into the Unit 2 reload analysis and will be incorporated into the Unit 3 reload analysis for the next cycle (Spring 1997).In addition, when the setpoints are verified, they are still required to be reset to 1%(proposed SR 3.4.3.1).Thus, since the analysis still ensure that all limits are maintained even with the expanded tolerance, this change is considered acceptable. This change is also consistent with the BWR Standard Technical Specifications, NUREG 1433.L2 The time to reach NODE 4 (reactor depressurized to<105 psig, Cold Shutdown)has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in NODE 3 (Hot Shutdown)within 12 hours has been added (Reference Comment M4 above).These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.0 BFN-UNITS 1, 2, L 3 Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 (gi SPec ikcrc]ion (g 1.a.3 pp)>crab;I gg~3.R.9.k Mo 3,4,q.c.es Lz w3 Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 e into the primary containment irom unidentified sources shall not exceed In add tion, t e total reactor coolant system leakage into the primary containment shall not exceed sea.v.e l l.Reactor coolant system leakage shall be checked t e ai s li Jal 8 r ore t least once per~hours.b.Qo 3,Q,Q,Q Qi'+in~p JlCvia~haytime the reactor is in RUlf NODE, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged axe~ay 24-hour period in which the'eactor ia in the RUE NODS cep as e ne n 3.6.C.l.c below.C~LCo Z.g,q g During the first 24 hour in the RUE NODE following STAEHJP, an increase in reactor coolant leakage into the primary containment of.,>2 gpa is acceptable as long as the requirements of 3.6.C.l.a are met.Rl kg gmggq~BFH Unit 1 3.6/4.6-9 ANENOMNr N0.I g 7 i~AGE~~>>-~ 8PECi K<C$og Q(QEC 07 19S4 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the-succeeding 24 hours for the sump system or 72 hours for the air sampling system.2 With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.~<<3'us4g;~hon P,~+~~BP'r4 1575 Z,q,q The air sampling system may be removed from service for a L3 period of 4 hours for calibration, function testing, and maintenance without Ad~R+o~-providing a temporary monitor.'A+gCQHl<g~~~8 3.If the condition'n 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed'n~"~4'+W th COLD SHUTDOWN CONDITION within hours.3.6.D hRlig Q 1.When more than onc relief valve is known to bc failed, an orderly shutdown shall bc initiated and thc reactor depressurized to less than 105 psig within 24 hours.The relief~alves are not required to be OPERABLE in the COLD SHPTDO~.ONDITIOH.+ca fD 8I-iv R(g~e pl d JF<guircd'can a,g Add P" once'A'on oP ml Rch'on L 4.6.D l.Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.2.In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is loving from the valve.BFN Unit 1 3.6/4.6-10 AMENDMENT NL R l

UNIT 2 CURRENT TECHNICAL SPECIFICATION IVIARKUP PAGE OF~ ~~la'g j ar v~'.rmzv rv't'tv'A'~~~~s var vr@@sit~ava~'v'~~v:si~~~H.'l0 gv)gl aa IA~~'%'If'LW le (a~~~~~~~~~a~~~~~~~~~I~~~~,~~AI~'I e&~.a~.'\~~~~~~~~~~~~~~~~~~~~~II~~~~I'~~~~~~~~~~~\A~~~~AI~'~~~~~~~~~~~~I I~~~~4P l~I I'll,~~~~~'S I'~~~~I~~~~I't~~AI S~'4 C~4;o~S.V.9 DEC 0'?199'I 2~Anytime irradiated fuel is in the reactor vessel and reactor coolant teaperature is above 212 F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to bc inoperable for any reason, the reactor may reasfn in operation during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.2.Vith the air sampling system inoperable, grab samples shall be.obtained and analyzed at least once every 24 hours.S~>~S 7.]Rmnoe Fog~""~~S~ar Is~~.yS<~IS S'~clod The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout rovidi a tea ora monitor.L3 Adct.Dc7'Iod P, P jeep~'~ego~3.If the condition in 1 or 2 above cannot be metD L11 ,~<<~c.orderly shutdovn shall be]V(d eepu,~&A 4o~B.Z AdcP 2M D Ac7104 Q)(Igf Co<JiSr~3.6.D initiated and the reactor shall be placed i the COLD WK CONDITION vithfn ours~Po7 5//~why 4,oD3plZ3o& JQ gong and When more than one rcl e valve is knovn to be failed, an orderly shutdovn shall bc initiated and the reactor depressurfred to less than 105 psig vf@EX~4 hours The relief valves are not required to be OPERABLE in the COLD SHUTDOW COHDITIOlf ~>3~S7 I/iCA77ohJ p'o~CAAo1665 7o g/Q I~7~g qg~N<<<~5'Cc77og 4.6.D 1.Appro~tely one-half of all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating, cycle.All 13 valves vill have been checked or replaced upon the completion of every second cycle.In accordance vith Speci f ication 1.0.lS, each relief valve shall be manually opened until theraocouples and acoustic monitors dovnstream of the valve fndfcate steam is floving from the valve.BFH Unit 2 3.6/4.6-10 hMENDMBIT N.2 29:=A!."." 3

UNIT 3 CURRENT TECHNICAI SPECIFICATION MARKUP PAGE OF SA c.i<i 0'o AUG 26 1987 l.a.R ppl ifct fy,'tip/CO 3 Lco gg g.~~<2+3 Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 F e c r coo ant ge into thc primary containment from unidentified sources shall not 5 In addition, the total reactor coolant system leakage into the primary containment shall not exceed SR 3,~.w.t 1.Reactor coolant system lcakagc shall bc checked y t sum an ai sam li st re ord at east once per hours.b.y)s&>n%AC FfC Js04LS Anytime the reactor is in RUB mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm avcragcd in vhich the reactor is in the RUE mode cpt as c c 3.6.C.l.c below.Co During the first 24 hour in'he RUN mode follov STARTllP, an increase in reactor coolant leakage into thc primary containment of>2 gye is acceptable as long..as the requirements of 3$%1.a"are met.Pl)Rdd 4C0 3.q,'I,a BPS Unit 3 3.6/4.6-9 NatmNr gO.g o 8 PAGE 3 DEC 0 7'1994 20 R>D/I@ion A+Requircst Avion 8,1 Ac<i'aN C (t s4 Co~d4'i~)t 3.6.D.Anytime irradiated fuel is in thc reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systens shall be OPERABLE.From and after the date that onc of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for.the air sampling system.The air sampliag system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout rovidiag a temporary monitor.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in thc COLD S CONDITIO vithin ours~34 l SNYPo~4 fnndifjoee Ift/P honte md When more than re c valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurised to less than 105 psig within 24" hours.The relimf snLLyes are not required to be OPERABLE in the COLD SHUT1XNN CONDITION. 2.With the air sampliag system inoperable, grab samples shall bc obtained and analyscd at least once every 24 hours.5SLeSAF'eQon &4 Qhagu+8 P'nl Z.S TS Z.q,5~+hiSSccgon D kQLhiF<Qt'.tl4~Bo2"~~><~Qnn(i'h'on oF~)Rch'on 4.6 D.1.hpproxijnately one-halt of all relief valves shall be bench-checked or replaced with a benc~hecked valve each operating cycle.hll 13 valves vill hav been checked or replaced upon the completion of every secoad cycle.BFN unit 3 F44 ipse'cc4Ãon A~~+pcs W Bkd%575 3.9,'3 xn W>s',ekion 3.6/4.6-10 2.Ia accordaace vith Specificatioa 1.0.IR, each relief valve shall be manually opened until thermocouples aad acoustic monitors downstream of the valve indicate steam is flowing from the valve NENMBlT NL I 86 q IP

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE ADMINISTRATIVE A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The total LEAKAGE limit applies at any moment, to the previous 24 hours (not any future or past 24 hour period).This results in a"rolling average" covering"any.24-hour period." Therefore, changing"any" to"the previous" does not change any intent.In addition, the current provision (CTS 3.6.C.l.c), which allows an increase in reactor coolant leakage into the primary containment of)2 gpm during the first 24 hours in the RUN mode following STARTUP as long as unidentified leakage and total leakage limits are not exceeded, is encompassed by proposed LCO 3.4.4.d which allows the same.LCO 3.4.4.d is worded differently (i.e., a 2 gpm increase in unidentified leakage within the previous 24 hour period in MODE 1)but means the same.Since there is no"previous" 24 hour period until being in MODE 1 for 24 hours, this limit does not apply for the first 24 hours.These are editorial changes only and as such are considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE A new requirement has been added to preclude pressure boundary LEAKAGE.An applicable ACTION has also been added.This is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE CTS 3.6.C.3 requires an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION within 24 hours when certain conditions can not be met.Proposed Action C will require the plant be in MODE 3 in 12 hours and MODE 4 in 36 hours.The addition of this intermediate step to the COLD SHUTDOWN CONDITION is considerqd more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.M3 The proposed applicability of MODES 1, 2 and 3 is more restrictive than CTS 3.6.C.l.a applicability of"Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F." The Startup Mode will now include the mode switch position of"Refuel" when the head bolts are fully tensioned. The change eliminates the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist.Currently, CTS 1.0.H allows the plant to be considered in the SHUTDOWN CONDITION and in the Shutdown Mode with the mode switch in the Refuel position (and other positions are allowed while in the Shutdown Mode)as permitted by notes to that definition. The allowance to place the Mode Switch in other positions has been moved to Section 3.10, Special Operations and Section 3.3.2.1, Control Rod Block Instrumentation. Any technical changes to these allowances will be discussed in the Justification for Changes to these Sections.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Details of the methods for performing this Surveillance are relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs.PAGE~Or~BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE"Specific" Ll The total LEAKAGE allowed has been increased to 30 gpm.No applicable safety analysis assumes the total LEAKAGE limit.The limit considers RCS inventory makeup and drywell floor drain capacity.The new limit of 30 gpm is well within the capacity of the Control Rod Drive System pump and the RCIC System, and is well below the capacity of one drywell equipment drain or floor drain pump, which is used to pump the water out of the collecting sump.The collecting sumps can also accommodate this small additional leakage rate.L2 The Frequency has been changed from 4 hours to 12 hours, consistent with the allowance in Generic Letter 88-01, Supplement 1.The supplement allows the Frequency to be extended to shiftly, not to exceed 12 hours.Browns Ferry Technical Specifications currently define the frequency of shiftly as 12 hours, thus, this Frequency is adjusted to coincide with this.CTS do not provide a period of time to reduce leakage prior to initiating an orderly shutdown.Proposed ACTIONS A and B allow 4 hours to reduce LEAKAGE within limits prior to initiating a shutdown.This is reasonable since the total leakage limits are conservatively below the LEAKAGE that would constitute a critical crack size.The 4 hour completion time for ACTION B is reasonable to properly verify the source.of unidentified leakage before the reactor must be shutdown without unduly jeopardizing plant safety.The proposed changes are consistent with the BWR/4 Standard Technical Specifications, NUREG 1433.L4 The time allowed to shutdown the plant when the required actions are not met has been changed from"in the COLD SHUTDOWN CONDITION within 24 hours" to in MODE 3 (Hot Shutdown)in 12 hours and MODE 4 (Cold Shutdown)within 36 hours.The proposed 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.The additional 12 hours allowed to reach Mode 4 is offset by the safety benefit of being subcritical (MODE 3)in a shorter required time.0 BFN-UNITS 1, 2,&3 PAQF~OF~Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.4-RCS OPERATIONAL LEAKAGE L5 Proposed LCO 3.4.4, RCS Operational Leakage, will add an alternative to existing requirement in Specifications 3.6.C.l and 3.6.C.3 that a reactor shutdown be initiated if unidentified leakage increases at a rate of more than 2 gpm within a 24 hour period.Under proposed Required Action B.2, unidentified leakage that increases at a rate of more than 2 gpm within a 24 hour period will not require initiation of a reactor shutdown if it can be determined within 4 hours that the source of the unidentified leakage is not service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids.This alternative Required Action is acceptable because the low limit on the rate of increase of unidentified leakage was established as a method for early identification of Intergranular Stress Corrosion Cracking (IGSCC)in Type 304 and Type 316 austenitic stainless steel piping.IGSCC produces tight cracks and the small flow increase limit is capable of providing an early warning of such deterioration. Verification that the source of leakage is not Type 304 and Type 316 austenitic stainless steel eliminates IGSCC as a cause of leak.This significantly reduces concerns about crack instability and the rapid failure in the RCS boundary.Also, the unidentified LEAKAGE limit is still being maintained and will continue to limit the maximum unidentified LEAKAGE allowed.This change is consistent with NUREG-1433. BFN-UNITS 1, 2, 5 3 Revision 0

CURRENT TECHNI AL SPECIFICATION MARKUP

~8-4Co 3.9e4 QCT]ddt r90 B Pr'~ca hfoQ W ifcHonS g 8 c dry I~a y 3 Al Anytime irradiated fuel is in the reactor vessel and reactor coolant tempera e is above 12'F, oth the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 urs for the sump system or~~emu's for the air sampling system., e air s pling system ay e r oved fr m serv ce for a per d of 4 hours or cali ation, funct n tes ng, and ma ntenan e vit ut provi a tern ora monit r Zeu w , Rcg'on 8.I$6 QQq5 I I QKC 07 1994 2.With the air sampling system inoperable, grab samples shall be obtained and analyze at least once every~hours./2 3.AhogS C+o If the condition in 1 or Q2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION m tiorA r~~~iT~i<l2howrc one f.e.D vithin h s.3~Ll 1.When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor dcprcssurizcd to less than 105 psig vithin 24 hours.The relief valves are not required to bc OPERABLE in thc COLD SHUTDOWN.~NDITIQN. 4.6.D l.Approximately one-hal of all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating cycle.All 13 valves vill have been checked or replaced upon the completion of every second cycle.See Su~gp;~ye C4lnoIt.5~8F:~bt Bg5 5qcg og 2.In accordance vith Specification 1.0.NM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.BFN Unit 1 3.6/4.6-10 AMENDMNT NO.2 Z3 OF~ S IAblE 3.2.51RtNENIAI ION IIAI INN)ISIS i INIO OR tl 5 stas~2 f I ent 0 qupa ra n F lou Inl ator Smp fill te I iaer Swp Punp Out Rate i Set ints>20.1 ain.I c i ain A~ctl abl se ne r ctor cool I leakage.2.idered par of swp s tea.LAG>"t.S.a Floor Orain F lcm Intcgrator Swy Fill Rale Iiaer Sllql P~1 Out Rate Iiaa:r A.i Ia.8.9 in.I.Used to ghteoaine unidentifiable reactor cool anl leakage.2.Considered part of sunup systea.gLo p q,g b Or@sell Air Sampling L3 c~Oas Pari cu ate backgfoungde I4 C~4 I o NOISES: (I)Qicnever a system is required lo be operable, there shall be one operable systco ellher autaaatic or annual, or7~<T'f>~the acliun required in Section 3.6.C.2 shall be laken.(2)alt ate sy a to de ine the le f lao aanual syst rcby I lee betve swp pwp~tarts is eaka ou because olune o he s u be knae~(3)titan Ageipt of alum, imacdiate acti ulll be lakcn to conf lra lhc ala and assess t~poss y f Increas+leakage. BFN Unit I

Function IIIR IN ES I IABIE 1.2.E%ICY FQt ll tEAII EC N functional Iesl gg q g Calibralion 5 I S R 9.'f 5 J.5.fnstrmi~nchec ~F loor Orain Simp F lm Intciirator Air SuplinII Systea SR 3..5.'K Rale IQ PI~s~once th LS~2 hrS Fino Orain 4'ilg le n l Rath,1laars ~ohqel rail c e oo ra n tag c one qar bicycle y PlORS i,Banal 4 qgly~.7r)blr q,Z.E.~nfl)Ainn hfe65 Are 434 raised i n 4 J)c~tr~f$~gag~3 3~~adjt SR a.e s.z.~zlda~1.Functional tests shall be performed once per DEc, nicAAGA~l.0'J N26 1999 2.Functional tests shall be pcr ormed be ore eac startup with a required frequency not to exceed once per veek.3.This instrumentation is excepted from the functional test definition. The functional test vill consist of in)ecting a simulated electrical si al into th surement channel.4.ested~ing~ogic s tern cti 1 tee~.L,Ay 5.Refer to Table 4.1.B.6.e ic sys func onal tee s shaQ incl+e a cail,ibratkqn ange gcr o erati cycle f time clay rc e anKtimerif ncccs1a fo~propert f th tri s terna The functional test vill consist of verifying continuity across thc inhibit vith a voltohnm)eter. S.Instrument checks shall bc performed in accordance vith the definition of instrument check (sec Section 1.0, Definitions). hn instrument check is not applicablc to a particular setpoiat, such as Upscale, but.is a qualitative check that the instrument ie behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check ie included ia this table for convenience and to indicate that an instrument check vill be performed on the instrument. Instrument checks are not required vhen these instruments are not required to be OPERhBLE or are tripped.9.Calibration frequency shall bc once/year. 10.Deleted 11.Portion of the logic is functionally tested during outage only.12.The detector vill be inserted during each operatiag cycle*and thc proper amount of travel into the core verified.13.Functional test vill consist of applyiag simulated inputs (eee note 3)~Local alarm lights representing upscale and downscale tripe vill be verified, but no rod block vill be produced at this time.The inoperative trip vill be initiated to produce a rod block (SRM and IRM inoperative also bypassed vith the mode evitch in RUN).The functions that cannot be verified to produce a rod block directly vill be verified during the operating cycle.BFH Unit 1 3.2/4.2-59 mml)@sr HS.1 64 PAGE OF 0 UNIT 2 CURRENT TECHNICAL SP ECIFICATION MARKUP p>aa DEC 0 7 89'I poJa J,2t 2~g~l;~l;l,g L.LO i 384 P t'Tlob5 P+Q w.p sQ eA~)4rsoCd 8 hnytime irradiated fuel is in the reactor vessel and reactor coolant tern erature is above 212'F, both the sump and air samp ng systems shall bc OPERABLE.From and after thc a e t at one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for thc'ump system or~muse for the air sampling system.4p~:~2.AC.4-on Sl With the air sampling system inoperable, grab samples shall be obtained and analyze at least once every hours.e air sampling syst may be r oved fro scrvicc fo a per d of 4 h urs for cali ation, f ction test and ma tenance thout providi a tcmpor ry monitor.Ae.Tlsg s c+b.6.D 3~If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall bc initiated and the reactor shall be placed in e COLD WR COKDITIOI vithin hours.3'~+~~TIP ICATJON F'og, CJJhuMc~epnJ>@~~43 rm 7R<J~qy]oQ 1.When morc than onc relief valve is knovn to bc failed, an orderly shutdovn shall be initiated and the reactor dcpressurired to less than 105 psig, vithi~~hours~ The relief valves arc not required to be OPERABLZ in the COLD SHUTDOWNÃCOMITIOS.4.6.D fgo~$pfH79OCdh) Co+I s7lo 12, goal J oat 1.hpproxiaately one-half of all relief valves shal)be bench-checked or replaced vith a bench-checked valve each operating cycle.hll 13 valves vill have been checked or replaced upon the completion of every second cycle.2.In accordance vith Specification 1,O.W, each relief valve shall be manually opened until thcraocouples and acoustic monitors dovnstream of the valve indicate steam is f loving from the valve BFH Unit 2 3.6/4.6-10 AMENMENT NO.Z Z9 PAGE

TASIE 3.2.E INSTRIICNIA IIRT NNIfORS tEAXAGE I ORYKtt System 2 qu pment Ora n Fl ntegrator Swp ll Rate TIa>>r Swp.Pmp t Rate Tie>>r Set ints A>20.1 min.<13.1 min Action I.se o erm ne dent I f e r tor coolant leak@a.gAs 2.Cons red part of swpgystea. F loor Orain Floe Integrator Smp Fill Rate 1 Ia>>r Smp Pmp Out Rate Tie>>r>80.i in%.9 sin.I.Used to determine unidentifiable reactor coolant leakage.2.Considered part of swp system.L<0 Or@el I Air Sanpl ing LE 3 Xgverage background or Gas and Part culate W hl lu I Cl NOTES: (I)lkenever a system'Is required to be operable, there shall be one operable system either autanatlc or manual, I rT><<or the action required In Section 3.6.C.2 shall be taken.A 2)An alte te system to determine the leakage f1~~a manual system<>her~the tie>>betuee~wp pwp tarts ls amitore~The tie>>interval<i&i determine the Ieakag~ou because the~m>>of s vill be knam.{3)Upo~ecelpt of alarm,<<n>>te action uill be taken to conf I@a the alarm and assess the Ibilit f increaL~leakage. BFNMlt 2 Function~TABIE i.2.~ININll'I@1 ANO CAI IBNAI ION FNE NCV tN ORAKLL LEAK OEIECIIQI INSIN~IAI ION sg 3.4s'.)Floor Oraln Sup Flee Integrator Air Sanpllng Systea SR 3..5.>te 3)A5 once ths eaealday L~(g.~LA4 I~a n S te and ce/r c~Ib9L fiber Oralnc'g.Aq I ance/operant cycle 8FNZJnlt 2

W'rg/4 aU<o qpc/-A'~4 0'2 g 7A r~m,~>ivy/a am<~raSs4,>>/4~~~~~yC<kCA~ZQ~)~~~~a/5$3.45.z.Fir~l.Functional tests shall be performed once per P~<<l ical(o'e 3 V.S'JAN 26 1999 20 Functional tests sha e per orme e ore eac startup with a require frequency not to exceed once per week.3~4~This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measur ed during logic stem functiiial teh4s.5.Refer to Table 4.1.B og c system unc onal te~ts shall include a o~ibration~nce ~er ope ing cycle o~e delay relays and timers necessary for proper.functio ng of the tri s stems 7.8.The functional test will consist of verifying continuity across the inhibit with a volt-ohmmeter. Instrument checks shall be performed in accordance with the definition of instrument check (see Section 1.0, Definitions). An instrument check is not applicable to a particular setpoint, such as Upscale, but is a qualitative check that the instrument is behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check is included in'this table for convenience and to indicate that an instrument check will be performed on the instrument. Instrument checks are not required when these instruments are not required to be OPERABLE or are tripped.9.10.Calibration frequency shall be once/year. Deleted Portion of the logic is functionally tested during outage only.12.The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.13.Functional test will consist of applying simulated inputs (see note 3).Local alarm lights representing upscale and downscale trips will be verified, but no rod block will be produced at this time.The inoperative trip will be initiated to produce a rod block (SRM and IRK inoperative also bypassed with the mode switch in RUN).The function that cannot be verified to produce a rod block directly will be verified during the operating cycle.~<<~sf'Pcr4i< 4r FA~qz~~8/575'3, BFN Unit 2 3.2/4.2-59 AMENDMECRV. y jy rAGF~~F~ UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF~ DEC 07 1994/7lpgg5,\2$3 2~t~Xg.4$(.f LOA 5 P~6 Pcc posed blok K Ac.A 5+8 Anytime irradiated fuel is in the reactor vessel and reactor coolant tern erat re is above 212'F oth the sump an air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or i~oars'or the air sampling system.Req~'nd@thon 8.I 3odAyp 2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every~hours., 12 Th aa.sam ng sys em ma be rem ved from service or a peri of 4 h s for calib tion, f ction te ting and mai tenance ithout rovidin a tern o moni or.3~/}cTlo gg c+u.6 D.BEN Unit 3 If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in he COLD SKJTDOMN CONDITION within ours e Ll When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurised to less than 105 psig within 24 hours.The relief'~ves are not required to be OPERABLE in the COLD SE/TDOMN CONDITION><>>'@AH'o g<Hhsb aP T~ggA/lSTS+hi5 Se~bo 3.6/4.6-10 ,~c Abr SlivgeWe CoAoi5ou>~l2.hours and.6.D.1.Approximately one-hal of all relief valves shall be bench-checked or replaced with a bench~hecked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.2.In accordance with Specification 1.0.NM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valvr NeoMENNO.Z 86 i AcE~oF~' 0 'ThNE 3.2.E TI5TRNKNT THAT ICNTTNS LfhlNiif TIITO tl~iov~aor~4r Rap Out Rate llmr gO.I ale.~i<<T3.4 ale.Used to dateaine Identifiable reactor ant Toalraoe.2.Considered of susp systea.Floor Oraln Flee lnte9rator LCO Susp Fill ILate Tlmr~'"'~I S~FLep Out gaQ Tieer LC.>3i'I.>ibOryuel 1 Alr SapllnII 0'Oa tart cu ate l.Used to date>aine unidentifiable reacior coolant leakage.2.Considered part of susp systea.RIG,: (1)lt>>never a systea ls rawlred to be terable, Mere sl>>II be one cperable systea eitber autcaatlc or aanual, or the action required le fectlon 3.C.CYshall be tahse.{2)An 1 terna ystea deterai the Teakyg~as ls a 1 s a+i ed.T Im Ia al uil detaealee 1 Tat 0>>mba susp s ts ls~J vo suep 1 be{3~rece of al Ieaedla tlon 1 be to coe the a aed sess poss lllty of rea sad e late A Q b

N I 1 Chil Yl F hNE.2.f CB-.floor Orala Sap fly'y4eyrator hir Qep I lac Systea 5g 3.,g.g I2)f5.liS Kqu t la F ala t c KN-lhlt 8 8)CL)CÃl

JAN 26 tS89 Only Noes l,g~d 0 aPP/9++$lbk q.2.E~r<~',.SPeC'afio~ 9..5~o<S~adcb<SccL i n~~r/MALS 6 San+ion p,3~s~tg+55y. 9J Sg, 3ego 5eX day l.Functional tests shall be performed once per 5 Functional tes s e 1 be performed before each startup with a require frequency not to exceed once per veek.3.Thiy instrumentation ie excepted from the functional test definition. The functional test vill consist of injecting a simulated electrical signal into the measurement channel 4.Test dur g lo 5.Refer to Table 4.1.B.al este.J R~J 6.e logic ystem functi 1 tests ha c u e ca b tion~ce ger op ating c le o time de y rela and t ers n cesar for prier)func oaing o he t e st 7.The functional teat vill consist of verifying continuity across thc inhibit wi,th a volt ohmmeter.8.Instrument checks shall be performed in accordance with the definition of instrument check (ece Section 1.0, Definitions). An instrument check is not applicable to a pareicular setpoint, such ae Upscale, but is a qualitative check that the instrument is behaving and/or indicating in an acceptable manner for the particular plant condition. Instrument check is included in this table for convenience and to indicate that an instrument check will be performed on the instrument. Instrument checks arc not required when these instruments are not required to bc operable or are tripped.9.Calibration frequency shall be once/year. 10.(DELETED)11.Portion of the logic is functionally tested during outage only.12.Thc detector vill bc inserted during each operating cycle and thc proper amount of travel into the core verified.C 13.Functional test vill consist of applying simulated inputs (see note 3).Local alarm lights representing upscale and downscale trips will bc verified, but no rod block vill be produced at this time.The inoperative trip vill be initiated to produce a rod block (SRM and IRM inoperative also bypassed with thc mode svitch in RUN).The functions that cannot be verified to produce a rod block directly vill be verified during the operating cycle.See Pug>'iak'on &r<tony S Gpss+5+BFN Unit 3 3'/4.2-58 AMENOMERF WP.y P g pAGE~o"~ 0 t JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS LEAKAGE DETECTION INSTRUMENTATION ADMINISTRATIVE Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BMR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 The revised presentation of actions is proposed to explicitly identify that LCO 3;0.3 is required to be entered if all.required RCS leakage monitoring systems are inoperable. This action is consistent with the current requirements and is considered a presentation preference. Therefore, this change is considered administrative. A3 The Table format is being deleted.This change is considered a presentation prefer ence.Therefore, this change is considered administrative. A4 Proposed ACTION B is modified by a note that explicitly states that the provisions of 3.0.4 are not applicable. This explicitly allows a mode change when both the particulate and gaseous primary containment monitoring channels are inoperable. This allowance is provided because, in this Condition, the drywell sump monitoring system will be available to monitor RCS leakage and the compensatory actions for the inoperable system will provide additional indication of RCS leakage.This is an administrative change since existing Technical Specifications do not have an explicit requirement that prohibits entry into a Mode or condition when an LCO required by that Mode or condition is not satisfied. Therefore, CTS allows the actions being permitted by the note being added.This is consistent with NUREG-1433, BFN-UNITS 1, 2, 5 3 Revision 0 A5 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION Frequency has been editorially changed from monthly to every 31 days and from every six months to every 184 days.This is an administrative change since these are equivalent time periods.A6 The current provision (CTS 3.6.C.2, 2nd paragraph) that allows the air sampling system to be removed from service for a period of 4 hours for calibration, functional testing, and maintenance without proyiding a temporary monitor has been eliminated. There is currently no requirement for a monitor for at least 24 hours (CTS 4.6.C.2).Therefore, the current provision serves no purpose.TECHNICAL CHANGE-NORE RESTRICTIVE The proposed applicability of NODES 1, 2 and 3 is more restrictive than CTS 3.6.C.l.a applicability of"Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F." The Startup Node will now include the mode switch position of"Refuel" when the head bolts are fully tensioned. The change eliminates the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist.Currently, CTS 1.0.H allows the plant to be considered in the SHUTDOWN CONDITION and in the Shutdown Mode with the mode switch in the Refuel position (and other positions are allowed while in the Shutdown Mode)as permitted by notes to that definition. The allowance to place the Mode Switch in other positions has been moved to Section 3.10, Special Operations and Section 3.3.2.1, Control Rod Block Instrumentation. Any technical changes to these allowances will be discussed in the Justification for Changes to these Sections.M2 The frequency of grab sampling with the air sampling system inoperable has been increased from 24 hours to 12 hours.A grab sample once/12 hours provides adequate information to detect leakage during the extended (See Justification for Change L4)period of time that the air sampling system is allowed to be inoperable. H3 Not used.M4 Not used.H5 The Frequency of the channel check requirement has been changed from every 24 hours to every 12 hours, consistent with Generic Letter 88-01, Supplement 1 and NUREG-1433. This is an additional restriction on plant'peration. BFN-UNITS 1, 2, 8L 3 Revision 0 ' JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION CTS 3.6.C.3 requires an orderly shutdown be initiated and the reactor to be in the COLD SHUTDOWN CONDITION within 24 hours when certain conditions can not be met.Proposed Action C will require the plant be in MODE 3 in 12 hours and MODE 4 in 36 hours.The addition of this intermediate step to the COLD SHUTDOWN CONDITION is considered more restrictive since CTS does not require any action to have taken place within 12 hours.The allowed Completion Time is reasonable, based on operating expe}ience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LA1 The description of an acceptable alternate system to measure leakage has been relocated to the Bases or procedures that support compliance with the limits for RCS Operational Leakage in proposed Specification 3.4.4.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures and FSAR will be controlled by the provisions of 10 CFR 50.59.LA2 The details relating to the setpoints have been relocated to the procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA3 The details relating to actions required upon receipt of an alarm have been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA4 Details of the specifics of the functional, calibration, and logic system functional test related to the floor drain sump fill rate and pump out timers has been relocated to procedures since the operability of the system is not dependent upon these timers.Changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 5.3 PAQE 0F Revision 0

LA5 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION The drywell equipment drain sump monitoring system functions to quantify identified leakage.Since the purpose of this specification is to provide early indication of unidentified RCS leakage, the drywell equipment drain sump monitoring system has been relocated to the Bases or procedures that support compliance with the limits for RCS Operational Leakage in proposed Specification 3.4.4.The design features and system operation are also described in'the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures and FSAR will be controlled by the provisions of 10CFR50.59. TECHNICAL CHANGE-LESS RESTRICTIVE"Specific" Ll L2 The time allowed to shutdown the plant when the required actions are not met has been changed from"in the COLD SHUTDOWN CONDITION within 24 hours" to in MODE 3 (Hot Shutdown)in 12 hours and MODE 4 (Cold Shutdown)within 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added.These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.This requirement has been deleted.An instrument check would not consistently demonstrate operability since normally the instruments could not be compared to any other instruments, and their reading could be anywhere on scale;thus, observing the meter would provide no valid information as to whether the instrument is OPERABLE.The CHANNEL FUNCTIONAL TEST requirement is the best indicator of OPERABILITY while operating, and this requirement is being maintained. This is also consistent with the BWR Standard Technical Specification, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.4.5-RCS LEAKAGE DETECTION INSTRUMENTATION L3 CTS Table 3.2.E defines the air sampling system as consisting of gas and particulate monitoring channels (i.e., both channels are required OPERABLE for the air sampling system to be considered OPERABLE). .Proposed LCO 3.4.5.b requires either one channel of the gas or one channel of the particulate monitoring system to be OPERABLE.This is less restrictive than CTS requirements but is acceptable since either channel is capable of indicating increased LEAKAGE rates thaf correlate to radioactivity levels of 3 times average background. L4 The allowed outage time for the air sampling system has been changed from 72 hours to 30 days.The 30 day allowed outage time recognizes that at least one other form of leak detection is available (sump monitoring) and takes credit for the increased sampling frequency of 12 hours (versus CTS of 24 hrs).This change is consistent with NUREG-1433.L5 The calibration frequency has been changed once pe}3 months to once per 18 months.This new Frequency is consistent with BFN setpoint methodology, which considers the magnitude of the equipment drift in the setpoint analysis over an 18 month calibration interval.The primary containment leak detection noble gas and particulate monitor is a digital Eberline continuous air monitor (CAM)which is identical to the building effluent monitors whose calibration frequency is 18 months in accordance with the Offsite Dose Calculation Manual (ODCN)and previously required by Technical Specification Table 4.2.K until these instruments were removed by Amendment No.216 dated September 22, 1993 (reference TS 301).Excessive calibration can cause damage to the equipment. In addition, plant operations could be impacted while the equipment is removed from service for calibration since it would not be available for leak detection. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE S GF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP clPi'z.V.4 J N2819%3.6.8.4~When the reactor is not pressurized vith fuel in the reactor vessel, except during the STARHJP COHDITIOR, the reactor vater shall be maintained vithin the folloving liaita.4.6.B.4~Whenever the reactor is not pressurized vith fuel in the reactor vessel, a ,saaple of the reactor coolant shall be analyzed at least every 96 hours for conductivity, chloride ion content and pH.ao Conductivity-10 yeho/ca at 25 C b.Chloride-0'ppa c.pH shall be betveen 5.3 and 8.6.Sec Sufkif eeHon Ar CJ4nst 5~'<<><8/9<8.ta this Sec gioa llnK A~sR3'.v,q,l 5~LCO 6.P 9.4 Ryl'c When the tiae liaita or maxiinm conductivity or chloride concentration liaita are exceeded, an orderly ahutdovn shall be initiated iaaediately. The reactor shall be brought to the COLD SHUTDOWN CORDITIOR aa rapidly aa cooldovn rate eraita.Wheneve the reactor ia r t cal the ta on activity concentrations in the reactor coolant shall not exceed the equilibritm value of 3.2 pCi/ga of dose equivalent I-131.CQuirc During guilibriaa pover operati an iaotop c ana aia nc y q t t tive ur cata for at 1 aat I 31-132 I-I-13 be perfoza eel~on a coolan iqaid aaaple.da)s A.+s.l k4ditional coolant samples shall be taken vhenever the reactor ctivi ceeda pre to equilibritm concentration specified in 3.6.B.6 c ti re t: BFR Qnit 1 3~6/4~6-7 AMENOMENT Ra.2 0 8 8~' ympmg N A.+ikey<<,'~s 4r Cu d.h'hA pcs(od@Tfog'8 This limit ma bc exceeded for a maximum of 4$hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gm encver e reac or s cr t cal s pcr tc mor 5X its earl po er op rati n er s cepti n fo thc li activi limits If the iodine concentratioa in the coolant exceeds 26 pCi/gm, the reactor shall be shut dovn, and the steam line isolation valves LQ, ufik4n tQ Qurg or be fn NM'I i0hi~34 h~rZ~f a.During the ST C HDITIOI b.ollovi a si fican over e40 c.,F lloving an,incre c in the equ librium of gas le 1 exceed ng 10, 0 pCi/ec (at the ste get ai e)ector)vithin a 48-hour period d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be takea at 4 hour interval or our, or ti a sta le iodine oncen rati a be ov the limit valu (3.pC ga)i establ shed.Hov cr, a least cons cutive samples shall b akea in al cases.kn isotopic analysis 11 be performed for each sample, and quantitative measuremcnts made to determine the dose equivalent I-131 concentratioa. 7.When there ia no fuel ia the reactor vessel, technical specification reactor coolant chemistry limits do not apply.7.When there is no fuel in the reactor vessel, sampling of reactor coolant chemistry at technical specification frequency is not required.**or the p rpose o this ection sampl frequ cy, a s ficant over chang>>is de ed as a change ceasel 15X f rated p ver ia ess t 1 hour.BFI Unit 1 3.6/4.6-8 NENOMENT lE 2 9 8 PAGE 3 QF~ 0 INSERT PROPOSED NEW SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. PAGE~OF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.4.7 RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN ECHNIC L C GE-0 ST IC IV Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 H(INSERT PROPOSED NEW SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications.

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTBI-COLD SHUTDOWN TECHNICAL CHANGE-ORE ES CTIV Hl A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent. with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS I, 2, 5 3 Revision 0 I 0 UNIT2 CURRENT TECHNICAL SPECIFICATION MARKUP JUNP,8 tsar 3.6.B.Coo 4.6.B.Coo e t@CO 34@hppl c,.4.When the reactor is not pressurized vith fuel in the reactor vessel, except during the STARTUP COHDITIOH, thc reactor vater shall be maintained vithin the folloving limits.a.Conductivity-10 pmho/cm at, 25 C b.Chloride-0.5 ppm c.pH shall bc betveen 5.3 and 8.6.5.When thc time limits or maxilla conductivity or chloride concentration limits are exceeded, an orderly shutdovn shall be initiated immediately. The reactor shall be brought to thc COLD SHUTDOMH COHDITIOH aa rapidly as cooldovn rate crmits.~t 6.Whenever thc reactor i rit ca thc limits on activity concentrations in the reactor coolant shall not exceed thc equilibrium value of 3,2 pCi/gm of dose equivalent I-131.4.Whenever the reactor is not pressurized vith fuel in thc reactor vessel, a sample of the reactor coolant shall bc analyzed at le'ast.every 96 hours for conductivity, chlori ion content and pH.5ec'X<sf<gi'c bio~+<+~d~far<TS 3.C.8/O'.C.g His S~4<o~4r$~3 H.9 I sR 3.'A6 I 5.Dur quilibrium pover o cratio an iaotop c analysis c itativc me urgents for at cast I-13K, I-132-133 and I-1 s be performed on a coolan liquid sample.H3 9 Joys LI j~.+'5, 6.Add tional coolant samples ahall be taken vhcnever the reactor activity exceeds cent equilibrium concentration 4.Al specified in'3.6.B.6 one~f o o c iti are+t: PAGE BFH Unit 2 3.6/4+6-7 NENOMENT RtL Z 2g o 0 (Q S cc,S;,4)..~ ~.q(JUN 8 8$994 p Cl'lo fJ p ETIO+0 Pygmy>gQ/vog~P kfNet'4J Ac~gQ,h c This limit may be exceeded 0 for L.of 48 hours.During this Lctivity transient the iodine concentrations shall not exceed 26 pCi/gm enever t e reactor is critical.e eac operated more t 5X of s year pover eration er s excep ion for e e uil brium ctivit limits If the iodine concentration in the coolant.exceeds 26 pCi/gm, the reactor Shall be shut dona, and the steam line isolation vLlves c W.~(2.w~goal Pep, A.t a.During the STARTUP COHDITIOH Folloving i4fgnffic pover change+c.olloving an in ease i the equilibri off as level excee ing 10,0 pCi/sec (at the steam et Lir e)ector)vi in a 48-hour eriod.d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hour intervals for 48 ours, or unt 1 o~be I'~AW Q~i&5$4v~a sta e iodine c entration belo the limiting va e (357, yCi/gm is established. Hovever, least 3 consecutive samples sha 1 cases ka isotopic analysis Shall be performed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concentration. Shen there is no fuel in the reactor vessel, technical specification reactor coolant chemistry lild.ts do not apply.7.When there is no fuel in the reactor vessel, sampling of reactor coolant chemistry at technical specification frequency is not required.For the purpose of this section sampling frequ, a si ficant poser e i defin as L change ceding 1SX of r ed pover in less than 1 hour.3'/4.6-8 AMEMOMENt'Mt. M M M PAGE

INSERT PROPOSED NEM SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. PAGE OF JUSTIFICATiON FOR CHANGES BFN ISTS 3.4.7 RHR SHUTDOWN COOLING SYSTEN-HOT SHUTDOWN T C NIC L C GE-0 EST IC IV Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BMR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0

INSERT PROPOSED NEM SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Speci fi cati ons.

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTEN-COLD SHUTDOWN TECHNICAL CHANGE-NORE RESTRIC IVE Hl A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in NODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2,&3 Revision 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP

3.6.B.4.6.B.C o 4.Mxen thc reactor is not pressurized vith fuel in the reactor vcsscl, except during the SThRTUP COHDITIOS, thc reactor vater shall be maintained vithin the folloving limits.a CoILductivity 10 pmho/cm at 25~C b.Chloride-0.5 ppa c.pH shall be betveen 5.3 and 8.6.4.Whenever thc reactor is not pressurized vith fuel in thc reactor vessel, a sample of the reactor coolant shall bc analyzed at least every 96'hours for conductivity, chloride ion content and pH.~><5+i~i~on Qe Changes'CI'~F 4 8/y.C..8;< +h;s Sccdon Ilfok r 5 3 Q.g, I 5.When the time limits or mm~mL conductivity or chloride concentration limits are exceeded, an orderly shutdovn shall be initiated immediately. The reactor shall bc brought to the COLD SHUTDOWN CONDITION as rapidly as cooldovn rate permits.(6.Whenever the I ritic the limits on activity 3,q,p concentrations in the reactor , coolant shall not cxcecd the (hyllcrk'IkPOoflibrf a ooloe of 3.2 PCi/ga of dose equivalent I-131.5~6~During qu libriua povc peration an sotopic aILa s s 2 tit tive eas cm ts fo at 1 ast-131 I-32, I-1 I-1 4 s bc performed oIL a coolant 1 quid sample.7$/5 e'.Ron,l 8.1 0 coo ant samples shall be taken vhenever the reactor activity exceeds 0 e en 0 e equ br ua concentration specified in 3 o th fo ovi c i i a m BFR Unit 3 3'/4'-7 AMENDMENT NL y 8 y F~GF~OF

n3.%6 jets)sg Qhon P (I+fon 8lo gpooS A4 oo gcpw o',CcP j44s'y~S 4w (e J4 A This liuent Iaay bc exceeded for a aaxiam of 48 hours.During this activity transient thc iodine concentrations shall not exceed 26 pCi/eaevcr reactor s cr tical e e ore o rate r than 5 of i s yea y vc opcrat oa er s e ion, for c c ilib a iv ty liaits If the i dine concentra on in the coolant exceeds 26 pCi/ga, thc reactor ahall bc shut down, and the steaa line isolation valves shall be clos bligh'on~ghmr or be on Q~9 lPo'Hlo&3Q~Op AC), At'H n k.l a.Duri the ST CO ITIOH b.Follovi a signi f ant pover ec*C~llovtng an increase the equili ri~of gas level ceding 10, 0 pCi/sec at the ste 5et air C5 tor)vithia a 48-hour riod.d.Whenever the equilibrhm iodine limit specified in 3.6:B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hour intcrv 1 o 48 ho s, r until stab iod coac trc ioa b lov liat v e.2 p/gR)s cata ished Ho ver, t least 3 c cu s ess lbet al cases isotopic analysis be pcrforaed for each saaple, snd quantitative aeasureaents sade to detcraine the dose equivalent I-131 concentration. 7.When there is no fuel in the reactor vessel, technical specification reactor coolant cheaistry limits do not apply.g~>WAX aNon 4 r C~g'>l'i~B (~oo,;s rC~7.When there is no fuel in the reactor vessel, saapling of reactor coolant cheaistry at technical specification frequency is not required.For the urpose f this section ssRpl f rcqu cyg a s fican power is dc cd as a change ccd 15X f rated vcr jn css 1 hour.BFI Unit 3 3.6/4.6-8 AMENOMgg~L Z 81 Gp g 0

INSERT PROPOSED NEW SPECIFICATION

3.4.7 Insert

new Specification 3.4.7, Residual Heat Removal System-Hot Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. ~~CE OF JUSTIFICATION FOR CHANGES BFN ISTS 3.4.7 RHR SHUTGOMN COOLING SYSTEM-HOT SHUTDOWN EC NICA CHANGE-0 E EST IC I Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS I, 2, 5 3 Revision 0

y(INSERT PROPOSED NEM SPECIFICATION

3.4.8 Insert

new Specification 3.4.8, Residual Heat Removal System-Cold Shutdown, as shown in the BFN Unit 2 Improved Technical Specifications. JUSTIFICATION FOR CHANGES BFN ISTS 3.4.8 RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN ECHNICA CHANGE-ORE ESTRIC IVE Ml A new Specification is being added requiring two RHR Shutdown Cooling subsystems to be OPERABLE in MODE 4.Appropriate ACTIONS and a Surveillance Requirement are also added.This is consistent.with the BWR Standard Technical Specification, NUREG 1433 and is an additional restriction on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 OF

13USTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY ADMINISTRATIVE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical, Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.~A2 Note is added to the Required Actions for Condition A to indicate that LCO 3.0.4 is not applicable. Entry into the Applicable Modes should not be restricted since the most likely response to the condition is restoration of compliance within the allowed 48 hours.Further, since the LCO limits assure the dose due to a LOCA would be a small fraction of the 10 CFR 100 limit, operation during the allowed time frame would not represent a significant impact to the health and safety of the public.In addition, this allowance is already inherently provided by the words of Specification 4.6.B.6.a, which states that additional samples are required"during startup" when specific activity exceeds the limit.Thus, this change is a presentation preference only and is considered administrative. A3 Existing Specification 3.6.B.6 requires that if the Dose Equivalent I-131 cannot be restored within 48 hours, or if an any time it exceeds 26 pCi/gm, the reactor must be shut down and all main steam lines must be isolated immediately. Proposed LCO 3.4.6, Condition B, allows the alternative of being in MODE 3 within 12 hours and Mode 4 within 36 hours under the same conditions. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).In Mode 4, the LCO requirements are no longer applicable. This change is considered administrative because existing 1.0.C.1 would require that the reactor be placed in Mode 4 within 36 BFN-UNITS 1, 2, 8L 3 Revision 0~u-;D JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY hours if the requirements in CTS 3.6.B.6 could not be met.This change is consistent with NUREG-1433. TECHNICAL CHANGE-MORE RESTRICTIVE Ml~e The Applicability has been changed to require the specific activity to be within limits in those conditions which represent a potential for release of significant quantities of radioactive coolant to the environment. Thus, MODE 3 with any steam line not isolated has been added.In addition, MODE 2 with any steam line not isolated has been added in lieu of MODE 2 when the reactor is critical.While this does allow the reactor to be critical with the main steam lines isolated while not requiring the LCO to be met, overall this change is considered more restrictive due to the MODE 2 subcritical and MODE 3 requirements. In addition, the ACTIONS have been modified to reflect the new Applicability, and an option for exiting the applicable MODES is-provided for cases where isolation is not desired.CTS 4.6.B.5 requires sampling reactor coolant to determine specific activity"during equilibrium power operation." Proposed SR 3.4.6.1, which contains proposed requirements for sampling reactor coolant to determine specific activity, is modified by a note that requires this Surveillance to be performed only in MODE 1.This change is slightly more restrictive because sampling will be required whenever the reactor is in MODE 1 and not just when equili6rium conditions have been established. This change is consistent with NUREG-1433.- M3 The Surveillance Frequency has been changed from monthly to weekly (every 7 days)for consistency with NUREG-1433, Rev.1.Since Revision 1 to the NUREG deleted the surveillance requirement to verify that reactor coolant gross specific activity is less than or equal to 100/E-bar pCi/gm every 7 days, the reactor coolant specific activity trending interval was decreased to 7 days from 31 days.TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" Revision 0 LAl CTS 4.6.B.6 contains requirements for reactor coolant and offgas system~~~sampling during startup, following significant power level changes, and BFN-UNITS 1, 2, 8L 3 2 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.6-RCS SPECIFIC ACTIVITY following significant changes in offgas radiation levels.The results of any of these samples are intended to determine if RCS specific activity is exceeding specified limits.Experience has determined that the weekly sampling required by proposed SR 3.4.6.1 and requirements for monitoring main steam line and offgas radiation levels is sufficient to ensure RCS specific activity levels are not exceeded.Therefore, RCS specific activity requirements for sampling stack gas, offgas and main steam line are being relocated to plant procedures and will be controlled in accordance with the licensee controlled programs.In addition, the criteria for when specific activity has been returned to limits (for 48 hours or until a stable iodine concentration below the limit has been established with at least 3 consecutive samples being taken in all cases)has been relocated to plant procedures and will be controlled by the licensee controlled programs.The method of determining dose equivalent I-131 (i.e., quantitative measurements of specific isotopes of Iodine), as described in CTS 4.6.8.5, has also been relocated to plant procedures. These changes are consistent with NUREG-1433.t"Specific" Ll Pro posed ACTION A allows the LCO limit to be exceeded for 48 hours provided that the specific activity does not exceed 26 pCi/gm.CTS 3.6.B.6 allows the limit to be exceeded during a power transient and limits the time the reactor can be operated, when the LCO RCS Specific Activity limit is exceeded, to less than 5%of its yearly power operation. Generic Letter 85-19,"Reporting Requirements on Primary Coolant Iodine Spikes," states that this limit is not necessary because reactor fuel has improved significantly since this requirement was established, and that proper fuel management by licensees and existing reporting requirements for fuel failures will preclude ever approaching this limit.Removal of this limit is consistent with the BWR/4 Standard Technical Specifications, NUREG-1433, requirements. L2 CTS 3.6.B.6 requires the reactor to be shut down and the'team line isolation valves to be closed immediately if the iodine concentration exceeds 26 pCi/gm.Proposed ACTION B allows 12 hours to close the isolation valves or to be in Mode 3.The 12 hour Completion Time is reasonable, based on operating experience, to isolate the main steam isola'tion valves, or to achieve the required plant conditions, in an orderly manner and without challenging.plant systems.The less restrictive 12 hour Completion Time is consistent with NUREG-1433. BFN-UNITS 1, 2,&3 Revision 0 PAGE~OF

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP PAG~ LI HG COHDITIOHS FOR OPERATIOH Rl SAd>Pi'yn 3.SURVEILLAH QUIREMEHTS 3.6 4.6 Applies to the opcrati status of the r actor coolant stem.Applies t the period cxaminatio and test rcquireacnt for the rca tor coolant syst To assure the tegrity and sa operation of the eactor coolan systems To etcrmine the ondition of the eactor coolan system and the o cration of th safety device related to it.Lgy>q,q 1.The average rate of reactor coolant temperature change during normal heatup or cooldowa shall not exceed 100'P/hr when averaged over a one-hour period.1.During heatups and s R i.q.q,)ooldovna the folloving parameters shall be recorded and ,reactor coolant temperature determined a minute intervals ti success rea at each given ocation are vi 5 P.a.Steam Dom Pressure (Convert o upp r vessel r gion temper ure)Reac r bot om drain tern eratur c.circul ion lo s A and B d.React r vcsse bottom head tempera re e.Rc ctor ve el shell a gaccnt shell flange BFH Unit 1 3'/4.6-1 PAGE OF A<2.Daring all operations vith a critical core, other than for lov-level physics tests, except vhen the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve 03 of Figure 3.6-1.2.Reactor vessel metal temperature at thc outside surface of thc bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to shell flange shall bc ecorde at least every minutes daring inservice bydrostptic or leak testing cn e vessel ressure is>312 psig.3.Daring heatup by nonnuclear means, except vhen the vessel is vented or as indicated in 3.6.k.4, daring cooldovn folloving~'"'l naclear shutdovn, or daring lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of carve 0?of Figure 3.6-1 until removing tension on the head stud bolts as specified in 3'.k.5~R2 3.est ecimcns repre cnt the reactor vcss, ba e veld and vel heat ffect xone met 1 be tailed in e r actor esse a scen to ves 1 v at the re mi plane 1 vel.The amber and e spe ens il b in accor ance GE repo t 1011.e spe ens hall ce the int t of ESTD 1-82.;.":-3 BFS Unit 1 3.6/4.6-2 IIIENDlNENT HO.17 0 SEP I 3 1995 4~L~o Z.9Q SR~~"fi1, NoQ Z, 5.sg 8.'f.$.5/V inc'2 The beltlinc region of reactor vessel temperatures during inscrvice hydrostatic )or leak testing shall be at or above the temperatures shovn on curve 41 of Figure 3.6-1.The applicability of this curve to these tests is extended to nonnuclear hcatup and ambient loss cooldovn associated vith these tests only if the heatup and cooldovn rates do not exceed 15 P per hour.Thc reactor vcsscl head bolting studs may bc partially tensioned (four sequences of the seating pass)provided the studs and flange materials are above 70 F.Before oading the flanges any more, thc vcsscl flange and head flange must bc greater than 80 P, and must remain" above 80'P vhilc under full tension.Nl~k, I P~pg~sw 3'.Q.5.2.S 0 z.9 f.g+8 a I<<~.s.9.v+Ivy 5'R 3oH.9 1+No4e 5.When the reactor vessel head bolting studs are tensioned and the reactor is in a cold condition, thc reactor vcsscl eratur inacdiately belov the head flange shall be fiopssrd 4<qgswlcs A r SRs 3.q,9.5)g+g BFS Unit 1 3.6/4.6-3 AMENDMBlT NO.2 P.f PAGE~OF~

S'kc'0'ca on 3.'f.9 6.~n gy,q The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant vithin the idle operati ecirc at, on loo s are vithin 50'F of each other.t SR Z.S8.S+~o<X 6.Prior o r R3 startu o an e recirculation loop, the temperature of the reactor i coolant n the o era and idle loops shal e en ly o e L,Co 7~The reactor recirculation pumps shall not be started unless the coolant temperatures betveen the dome and the bottom head drain are vithin 145 F.7.Prior to starting a+'recirculation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall be compared e o H3 P'.po'd Acti<~34s+;F;wg'<<gg, $r BFA 1575 3.q.~S~aMi'cz4on @r Cj~~ggc 4i BFQ tSvs z 4.6.E.~Jt~g~2.Whenever there is recirculation flov vith~the reactor in thc STARTUP or RUH Mode and one recirculation pump i is operating, the diffuser to lover plenum differential prcssure shall be checked daily and the differential pressure of an individual jet pump.in a looy shall'ot vary from the mean of all Jet pumy differential pressures in that loop by more than 10K.3.6.F 4.6.F The reactor shall not bc operated vith onc recirculation loop out of service for morc than 24 hours.With thc reactor operating, if onc recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWN COHDITIOH vithin 24 hours unless the looy is sooner returned to service.l.Recirculation pump speeds shall be checked and logge at least once pcr day.Sg 3Aegi Q 2~3~Folloving onc pump operation, the discharge valve of the lov speed pump may not be opened unless the speed of the faster pump is less thaa 5OX of its rated syeed.When the reactor is not in thc RUN mode, REACM POWER DPERATIOH vith both recirculation pumps out of-service for to 2 hours is yermitted During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is vithin 75'F of the saturatioa 2.Ho additional surveillance required.se~.~.v.v 3.Before starting cithcr recirculation pump during REACTOR PO OPERA the oo s rge temperature and dome saturation tempcraturc. BFH Unit 1 3.6/4.6-12 AMENDMEHT NP.2 y y pAGE~GP~ 0 3.6.F 3.9~9.'9~A/os 2 temperature of the reactor vessel vater as determined b dome pressure.e total elapsed time aa ural circulation aad oae pump operation must be no greater than 24 hours.S~e Y~swkc~g'on @kc 9t-Iv f 5 pg p,q,~The reactor shall not be operated vith both recirculation yumya out-of-service vhile the reactor'ia in the RUB mode.Folloving a trip of both recirculation pumps vhile in the RUN aode, immediately ate a manual reactor scram.3.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained ia accordance vith Syecificatioa 4.6.G throughout the life of the plant.a Mith ths structural integrity of any ESNE Code Class 1 equivalent coaponent, vhich is part of th>>primary system, not confozILing to the above requirements, restore the structural integrity of the affected component to vithin ita limit or maintain the reactor coolant system in either a Cold Shutdova condition or less than 50'F abide the minimum temperature required by HUT considerations, until each indicatioa of a defect has been investigated and evaluated. 4.6.G I l.Iaserrice inayectioa of MME Code Class 1, Class 2f aad Class 3 components shall be performed ia accordance vith Section XI of the ASIDE Boile aad Pressure Vessel Code aad applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except Mere syecific vritten relic haa baca graated by HRC yursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additioaal inayectiona shall be perforaed on certain circumferential yipe velda to provide additional protection against pipe vhip, vhich could damage auxiliary and control systems.5<<3KHiAczg'q ~*'TS S,S.g/q,g y BFS Unit 1 3.6/4.6-13 AMENOggPNg, p p 6;pA'8 OF~~

ClS gg JUN 2 8 1994 3.6.B.1.PRIOR TO ST TUP and at steaming rates'ess than 0,000 lb/hr, the folloving limits sh 1 apply.4.6.B.C o 1.Reactor coolant shall be continuously monitored for conductivity except vhen there is no fuel in the reactor vessel.a.Cond ctivity, pmh/cm at 25 C 2.0 b.oride, ppm 0.1 Whenever the continuous conductivity monitor is inoperable, a sample f reactor coolant hall be analyze for conduct vity every 4 hour except as liste belov.If the react r is in COLD SHUT OMH COHDITIOH, a s e of reactor co ant shall be lyzed for nductivity every hours.b.Once a veek the continuous monitor shall be checked vith an in-line flo cell.This in-line conductivity calibration hall be performed e ery 24 hours vhen ver the reactor c lant conductiv ty is>1.0 pmho/cm t 25 C.2.At steaming rates greater than 100,000 lb/hr, the folloving limits shall apply.a.Conductivity, pmho/cm at 25 C 1.0 2.During, star p prior to pressurizi the reactor above atm pheric pressure measurements of reac r vater quality shall performed to shov confo ance vith 3.6.B.1 of li iting conditions. b.Chloride, ppm 0.2 BFH Unit 1 3.6/4.6-5 AMENOMENT RO.2 0 8 FASP/OF~~ DEC 0 7 L994 3.6.B.4..B.Coo 3~ht steaming rates greater 100,000 lb/hr, thc reactor vatcr qua ity may cxcecd S cification 3.6.B.2 nly for the time 1 its specified belov.Exceeding these time imits or the follow ng max quality limits s all be use for placing the reactor in the CO SHUTDOMR CO ITIOH.a.Conductivity time abov 1 pmho/at 25'C-2 v ks/year.Maxim Limi t 10 mho/cm at 25'C 3.whenever thc reactor is operating (including HOT STAHDBY CORDITIOH) m suremcnts of reactor ter quality shall.be erformed according to the folloving schedule: a.Ch1oride ion content and'H sha be measured least once every 96 ours.b.Chlori e ion conte t shall be meas rcd at least eve 8 hours.vh ever reactor c ductivity is.0 pmho/cm t 25 C.b.Chlor de co entration time a ove 0.2 ppm-2 weeks/year. Maximum Limit-0~5 ppmo c.The reactor shal be placed in thc S WH COHDITIOR if pH<5.6 or>8.6 for a 24-hour period.c.h sample of actor coolant sha bc measured f pi at least onc every 8 hours vh ever the reactor oolant conduct ity i,s>1.0 pmho/at 25 C.BFH Unit 1 3.6/4.6-6 AMENOMEHT NIL R 1 3

~~.e.c'.e,s'ON28m 3.6.B.4.When t c reactor is not p csaurizcd vith fu in thc eactor vessel, cx pt dur thc SThRHJP CO ITIOH, the reactor vater s 1 be ma ntaincd vithin t f lloving limits.Conductivity 10 pmho/cm t 25'C 4.6.BE 4~cnevcr the eactor not ressurizcd ith fuel in the reactor csacl, sample of e react r coolant s ll bc lyzcd at least every 96 ours for conductivity, chloride ion content, and pH.b.Chloride 0.5 ppm c.pH sha be betvccn 5.3 8.6.5.When th time limits r conductivity or chlor de concentrat on limi s are cxceede , an ord ly shutdovn ll be in iatcd imacdia ely.The re ctor shall be brought to the COLD SHUTDOWNS COHDITIOH aa rapidly aa cooldovn rate ermita.5.During equilibrium pover operation an isotopic analysis, including quantitative miasurcmcnta for at least I-131, I-132, I-133, and I-134 shall be performed monthly on a coolant liquid sample.6.Mxcnever the reactor, is critical, thc limits on activity concentrations in the rcictor'oolant shall not exceed the equilibrium value of 3.2 pCi/gm of dose equivalent I-131.6.Additional coolant samples shall be taken vhencvcr thc reactor activity exceeds one percent of the equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are met: BFS Unit 1 3.6/4.6-7 AMENOMENT gg, p O 8 PAep 3.6.B 4.6 3.6.B.6 (Cont'd)This limit may be exceeded folloving pover transients for a maximum of 48 hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gm vhenever the reactor is critical.The reactor shall not be operated more than 5X of its yearly poser operation under this exception for the equilibrium activity limits.If the iodine concentration in the coolant exceeds 26@CD/gm, the reactor shall be shut down, and the steam line isolation valves 11 be closed immediately. 4.6.B.6 (Cont'd)a.During the SThEEP COHDITIOH b.Folloving a significant ponr change~+c.Folloving an increase in the equilibrium off-gas level exceeding 10,000 pCi/sec (at the steam jet air ejector)vithin a 48-hour period.d.Whenever the equilibrium iodine limit specified in 3.6.B.6 is exceeded.The additional coolant liquid samples shall be taken at 4 hou intervals for 48 hours, or until a stable iodine concentration helot the liNLiting value (3.2 pCi/ga)is established. Honver, at least 3 consecutive samples shall be taken in all cases.ka isotopic analysis shall be performed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concentrations 7.When ere i no fu he ,re tor v sel, chni 1 s ecifi tion r seto coolant eais ry 1 s do ot apply.7.Wh there no uel reacto vess 1)ling f rei tor olant eNList at t chni 1 specifi tion freq ency is not required.**For the purpose of this section on sampling frequency, a significant pover exchange>>is defined as a change excee~lng 15Z of rated pover in less than 1 hour BFH Unit 1 3.6/4.6-8 NENDMENT R5.Z 0 9 N Cl CTS p.6,g NY3>m 3.6.F 3.6.F.3 (Cont'd)temperature of thc reactor vessel water as determined by dome pressure.The total elapsed time in natural circulation and one pump operation must be no grcatcr than 24 hours.St.'c Wus~Flc<A'a~ g,(g)9 8F~ISTIC 3,1')Rc c'<<culm'ova ~Ps oPt<eh'~)i w+gs sccw>n 4.The reactor shall not be operated with both recirculation pumps outof-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G 4.6.Q 1.The st ctural integri of ASME Code C ss 1, 2, and equivalent compon ts shall be intaincd in acco cc with Spe fication 4.6.G thro out the lif of the plant.a.Mith the stra tural integrity of any ASME ode Class 1 equivalent omponent, which part of th primary system, ot conform to the above requirem ts, restore the structur 1 integrity of e affectc component to w hin its 1 t or maintain e react coolant syst in eithe a Cold Shutdo cond tion or less 50'F abov the minimum t peraturc required by HDT co iderations, until each indication of a defect has been invcstigated-and evaluated. Inservice inspection of AS Code Clas 1, Class 2;and Class 3 omponents shall be perform d in accordance with Sectio XZ of the ASME Boile and P ssure Vessel Code and appli able Addenda as rcqu red by 10 CPR 5 , Sec ion 50.55a(g), cept wh re specific wri ten relief s been granted HRC rsuant to 10 C 50;ection 50.55a((6)(i).Additional i cctions hall be performed n certa circumfercnt al pipe lds to provide add tional pr tection against pi whip, w ich could dama c auxili ry and control sy terna.BFK Unit 1 3.6/4.6-13 AMENDMQP'go, p p 6 PAGE 5 OF'

3.6.G 3.6.G.(Cont'd)b.Wl the struct al integrity o any ASME C e Class 2 3 quivalent omponent no conformi to the ab e requir ents, rest e the stru ural integ ty of the aff cted compo t to vithin i limit or solate the ffected co onent from all OPERAB systems.4.6.G.For Unit 1 an a ented inservice surveil ce program shall be peiformed to monitor pot tial'orrosive effect of chloride res ue released uring the March 22, 75 fire.The augmented ervice surveillance ogram ia specified as f lovs: a.Brows Ferry Me cal Ma tenance Instruct 53, date eptember 22, 19 5, paragra 4, defines the liquid p trant examinatio required during the f st, second, third and four refueling outages follovi the fire restoration. b.ma Ferry Mechanical Mai ce Instruction 46, dated ly 18, 1975, hppendix defines the liquid pens t examinations re uired during the sixth refueling outage folloving the fire.restoration. BFH Unit 1 3.6/4.6-14 AMENDIHBIT lS 80 6 PAGg JOL 0 s 8%3.6.H.ggg~During a modes of operatio , all snub crs shall be OPE except s noted in 3.6.8.1.All sa ty-related snubber arc 1 ted in Plant Surve llance Instructio l.ith onc or morc snubber(s) inope ble on a system that i required to be OPERABLE n the current plant ondition, vithin 72 ho s replace or restore e inoperable snubber(s) o OPERABLE status perform an engineer evaluation on the a tached compon t 02 decl rc thc attache system inoperable folio the appropri e Limit ng Condition statement for that system.4.6.8.~S Ea safety-related snubber s 11 be demonstrated 0 RABLE by performance f thc folloving a gmcnted inservice inspect on program and thc rcquir nts of Specificatioa .6.H/4.6.8. These snubber are listed in Plant Survci ance Instructions c 0 hs scd in this s cification,"typ of ubbcr" shall me snubbers of the c design and manu acturer, irrespective capacity.2~V Snubber are categorized as ina essiblc or acces ible during re ctor ope tion.Each o these ca egories (inacc ssible accessible) y be nspectcd inde cndently according to c schedule dctermincd Table 4.6.H-1.e visual inspecti interval for each t c of snubber shall be de crmined based on the riteria provid in Ta e 4.6.8-1 and e first i paction intcrv 1 termincd usi this criteria shal bc based upon the pr ious inspection nterval as establis d by the requir cnts in effect before amendment No.210 BFH Unit 1 3.6/4.6-15 NEMMENT IN, 2 go PAG~~ Ol 4.6 H mShhma 3~sual inspec ons shall verify that 1)the snubber has no vis le indications of damag or impaired OPBRhBI TT, (2)atta ents to th fo tion or su portiag st ture are ctional, (3)fast rs for the tachment o the snubber o the comp ent and to the snubber an orate are functio.Snubb rs which appear operabl as a result f visua inspe tions s 1 be cia ified cceptable and be reels ified acceptable or the purpose of establi ing the next visual i paction interval provided that (1)th cause of the e)ection is clearly estab shed and r edied for t particu r snubber for other s bbers espective o type that may be generi ally susceptible; and (2)the affected sn ber is functional y tested in the as-found ondition and determi d OPBRABLS per Specif ation 4.6.5.h revie and evalu tion shall be p rformed documented to)ustify co inued operation wi an unacceptab snubber.If continued peration c ot be gusti ied, the snu er shall declared inope ble and the MITIHG COHDI IOHS POR OP TIOH shall be met.BFH~Jnit 1 3.6/4.6-16 AMB1BEEtlTHg, 2 IP PAGE

C75 Z.b,h q,g, H 4.6,8~SggZZa 4.6.8.3 (Cont'd)hd itional , s bbers a tached sc ions of afety-re ated systems that ve cxp rien ed un ected potenti lly amagi transi ts ines t last insp tion eriod hall be eva ated for thc poa ibil y of c cealcd d e d func onally tested if appl cable, o confi OPBlhB ITf.Snub rs vhi have b cn mad inopcra e as re lt of expected tr ients isolate d e, o other r om events, hen thc rovisio of 4.6..7 and 4 6.8.8 e been m t and other appropriate co rectivc action implem ted, sh 11 not be counte in determining e next isual inspection interval.BFH Unit 1 3o6/4.6-17 .AMENOMENt'lN. 2 To 4.s ldll m ing ch refueling outage a represen tive sampl of 10K of e total of e ch type of saf ty-related ubbers in us in the pl t shall be f ctionally ested either in place or n a bench test The repre tative s le selected or functi 1 testing shall incl e the variou configura iona, opera ing enviro ents, and the ange of si e and ca city of s bbers vithln the types.e representat e sample should be ighed t include m e snubb rs from severe s ice ar as such as near cavy e ipment.The roke se ing and the secu ity of steners or attachment the sn hers to the corn onent an to the snubber chorage all be verifie on snubb rs select for FUH IOKLL TBSTS.BPH Unit 1 3.6/4.6-18 AMENDMENT NO.2 10 pAG a~OF~ Cl, 0, lAN 1g 1ggg 4.6.H.~S u i~<<i 5.C 0 C ter The s bber CT OHAL TEST hall" erif that: a.Activa ion restraining aetio)is chieved in b th t ion and corn ressi vithin the sp cified range, capt t t ine tia dep dent, celer ion lim ing echani al snubb rs may be tes ed to ve fy only at activ tion takes place in oth dire tions of ravel.b.Sn ber bleed or re ease vher required, i present i both c mpression and t ion ithin the pecifi d ange.c.For mech ical sn bbers, the for requir to initiat.or main ain motion of the s bber is not g at eno to overs ress the attached pipi or comp nent dur therma movement, or indica e impendi fa ure of e snubber.d.r"snubbe s specific lly equired ot to disp ace under co tinuous lo the abi, ity of the nubber to vit tand load ithout displa ement shal be, verified.BFH Unit 1 3.6/4.6-19

~6.8.4.6.8.5 (Con d)e.cating met ds may be used to mc ure parameters indircctl or parameters other th those specified if thos ,results can be correl ted to t e speci cd par eters thro estab ishcd me ds.6.kn cngin ring eva tion shall b made of ch failure to mee the FUR OKhL TEST accep ce crit ia to dete e thc ause of the fai ure.The result of this ysis s 1 be used, if plicable in select nubbers be teste in thc subscqu lot in effort to detcrai c the OPE ILITY of other ubbers v ch may bc sub)t to the e failure mod.Sclecti of snubbers fo future te ting may a o bc b scd on th failure lysis.or each s bber that does ot meet t e FUKCTIO TEST ace ptance criteri , an addi onal lot equal o 10 pere t of the rcaa cr of t t type of snu ers shall e functions ly te tcd.Tes ng shall ntinue un 1 no additi 1 noperable snubbers arc found vithin s sequent lots or all snubber of the orig 1 HJKCTI TEST typ have been teste or all susp ct snubbcrs iden fied by the failure ana sis have bc tested, as applicable. BFH Unit 1 3.6/4.6<<20 AMENDMgPNy ~>>PAGE~OF~

4.6.H.u b 4.6.H.6 (Co'd)any snubb selected or functio 1 testing either fai to lo up or fails move,.e., frozen i pl'ace, he cause vill be valuat and if caused y manu cturer or desig defici cy, all snub ers of e same des gn subj t to the s e defec shall be ctiona y tested is tes ing requi ement shall b independ nt of the r uirements stated abov for snubb rs not mee ng the CTIONAL TE accept e criteria.he discov of loose'r missi attachmen fastener vill be e luated to dete ine vheth r the cause ay be loc ized or gener c.The r ult of the valuation ill be use to sele other su pect snu ers for v rifying e attachme t stener , as applica le.7.S a be S Fo the snubber s)found i operable, an engineerin valuation sh ll be perf med on the compo ents which are restrained y the snub er(s).The purpo of this engineer g evaluati n shall to dete ine if the component, restra ed by the BFN Unit 1 3.6/4.6-21 AMENOMBfTNg f63

VAJA>ivs8~4.4.6.8.7 (Cont'd)8~snubber s)vere a ersely affect by the operability of th snubber(, and in orde to e sure that he restrained corn onent rem's capable of , m ting the signed service.u o a 0 S S b Snubbers hich fail he visual inspect on or the FUNCTI AL TEST ac eptance'rite a shall be repaired or r laced.Re lacement snu ers and sn bers which ha e repairs v ch might a fact the FUN TIONAL TEST esults shall meet the FUNCTIONAL ST criter before inst llation i the unit.The e snubbers shall have met e accepta ce criteria subsequent to their most re ent servic , and the FUNCTI NAL TEST m st have been erformed v hin 12 mont s before b ng installed in he unit.9.Permanent or other mptions from vis al inspecti ns and/or unctional t sting for i ividual sn bers may be gr nted by th Commission if justifiabl basis for ex ption is p sented and if applicable nubber life destructive sting vas perform to qualify snubber OP BILITY for the appli able design conditio at either the BFN Unit 1 3.6/4.6-22 AMENDMQP gP.y 8 g Cl 0, iJAN i 9 1989 4.6.H.4.6.8.9 ont'd)completi of thei fabrica ion or at subsequent date.Snubbers exempted shel continue o be liste in e plant i structions ith fo notes in cating the tent of t exemption 10.S e o a C The se ice life o snubbers may b extended b ed on an eva ation of th records of~FU TZONAL TEST m intenance hi tory, and nvironmenta conditions to which the s bbers have been expos d.BFN Unit 1 3.6/4.6-23 elemMM 50.18 8 0 0 JAN i9 1888 THIS PAGE INTENTIONALLY LEET BLANK BFH Unit 1 3.6/4.6-234 AMENOMENT NO.1,6 8 PAGE~OF~I W

~~~~~~~I~~~~W~~~~I~~~I~~~~~~t~I~~~~I I I~Il~g~~~.~~~'I~~~~~~~~~~~~~~~'~~~~I~I'~~~~~~I~~~~~~~~~I~~~~'~~~~~~~~,~~~~~~~~~~~~~~~'~~~~~~I AI~~~~~~~~~~~~~I~~~~~~~~~~~~~A~~~~~~~~~~~~~I~~~~: 'I~~I I I~'~~~I~~I~~~I~~~~~~~~~~~~~~~~'I~~~~~~I~~~~~~~I~~'~~I~~I,~~~~I'~~~*~~~~

Table 4.6.8-1 (Continued) JOt.0sm SHUBBER VISUAL IHSPECTIOK IHTERVAL Hote 4: If e number of una eptable sn hers is e 1 to or ess the nuaber in lpga B but rester the num r in Co A, the next pection i terval 1 be the arne as previous inte al.ote 5: f the number of unicceptab snubbers s equal t or, great than the number in Column , the next pecti interval shall be tvo-rds of th previous erval.ovever, i the number of una ceptable ubbers is 1 s than t e number Column C, bu greater the numb r in Col B, the ext interval s ll be red ced proporti lly by terpolat on, that is, e previo interval ll be red ced by a actor that is e-third the ratio o the diff ence be en the number o unaccep ble snubbers found dur the pr ious interva and the timber in Col B to differ e in the number in Col B and C.te 6: The provisions of Specification 1.0.are applicable for all inspection intervals up to and including 48 months.BFH Unit 1 3.6/4.6-23c AMENOMENT IL 2IO PAGE

t Section 3.4, Reactor Coolant System (RCS)Bases The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content of the proposed Browns Ferry Unit 2 Technical Specification Section 3.4, consistent with the BWR Standard Technical Specification, NUREG 1433.The revised Bases are as shown in the proposed Browns Ferry Unit 2 Technical Specification Bases.BFN-UNITS 1, 2, 5 3 Revision 0 pAGE~oF LI UNIT2 CURRENT TECHNICAL SPECIFICATIPN MARKUP Cl 0 LIMITIHt COHDITIOHS FOR OPZRATIOH P,l URVEILLAHCZ REOUIRZmHTS 3.6 S 0 4.6 S~S 0 cab Applies o the operating status of the rc tor coolant system.~0~v To assure the integ ity and safe operation of the rca or coolant system.Applies to the periodic examination an testing requirements for c reactor coolant system.~Oa~v determine thc condition of th reactor coolant system and thc operation of thc safety devices related to it.t 0 LCo 3.g,g 1.The average rate of reactor coolaat temperature change during normal heatup or cooldom shall not exceed 100 F/hr vhca averaged over a oae-hour period.sm 3.4.9.(1.)During heatups and followed~parameters shall be recorded and reactor coolant~pl temperature determined attic=minute intervals un s ve readings each given lo tion are vithia F.g.4 l a earn Dome Pre sure Convert to pcr vessel rcgi temperatur )Reacto bottom drain tcmpc ature Re rculation 1 ops B d Reactor ve sel bo tom head tern erature e.Rcacto vessel shell adjacent to shell flange BFE Unit 2 3.6/4.6-1 A>na~

2.During all operations with~o a critical core, other than for low-level physics tests, ezcept when the vessel is vented, the.reactor vessel shell and fluid temperatures shall be at or above the temperature of curve 03 of Pigure 3.6>>1.s Reactor vessel metal temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell ad/scent to shell flange shall be dc at least every~<'2 minutes during inscrvice hydrostat'ic or leak test hea the vessel pressure is>3L?psi@.3.Daring heatup by nonnuclear means, except when the vessel is vented Lco or as indicated in 3.6.k.4, during cooldown folloving nuclear shutdown, or during lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve 02 of Figure 3'-1 until removing tension on the head stud bolts as specified in'6.i 5~cst spec ens'eprcseat the reactor vessel, e veld, and veld he t affected zone metal hall be installed in e reactor vc scl a aceat to th vessel all at the re midplane level.Th number and type of ecimens vill be in cordance vith repor KDO-10115. e specimens shall m t the intent of ASIAN E 85-82.<<~Jf Unit 2 3 6/4 6-2 PAGE~AMENDMENT NO.17 0

5 c icAF>0 SEP i 3 1995 ss a 4~gCo g.A SR 3.g.9.I,~oR z.5.l QO 3qq The beltline region of reactor vessel temperatures during inservice hydrostatic) or leak testing shall bc at J or above thc temperatures shown on curve 41 of Figure 3.6-1.The applicability of this curve to these tests is extended to nonnuclear heatup and ambient loss cooldovn associated vith these tests only if the heatup and cooldovn rates do not exceed 15 F per houre The reactor vessel head bolting studs may be partially tensioned (four sequences of the seating pass)provided the studs and flange materials are above 70'F.Before loading the flanges any morc, the vessel flange and head flange must be greater than 82 F, and must remain above 82 F vhile under full tension.4~Si2>06 l No~]Hl p~,pop sg 5.4.9 2 Sg K.iJ.).S d Na4.I S'g?.q.'t-g +Uo~sg s.g.)7 p No+5.When the reactor vessel head bolting studs are tensioned and the reactor is in a cold condition, the reactor vessel shell temperature haaediatcly belov the head flange shall be P o~H C~p e<<dc5 gc.K~g.gq g (pg PAGE OP BFK Unit 2 3'/4.6-3 AMENDMENT NO.2 3 9

gy'le.~~~+It i f~~%t~~~~~~g~~~~'~4~~IAI~','i~eP a 0.TIVE~'%~~~A~'~LT~MWEW.~ XS,At~~~~A~~~~~~,~~o~f~~~~~: I~~~~~~~~~~~~~~~~~~~~~~.~~~~~~~~~~~~~~-'~~~~~~A~~~~A~'~~~~~IA~~I~~,~~~A~~~A~~0 I~~~~~I I~~~~A~~~~~~~~~~.~~~A~~~~~~ 0 ~AI 4.6.E.~Je RmSm~c4 Gw~ifi<4ia~ waar 6"~~gu 4~[gl57$3,'I 2 See Z~S]"Ia.h'~4-Ct~)~8~~Isis Z.q.i 2~Whenever there is recirculation flov vith the reactor in thc SThRTUP or RUH Node and one recirculation pump is operating, thc diffuser to lover plenum differential pressure shall be checked daily and the differential pressure of an individual get pump in a loop shall not vary from the mean of all get pump differential yrcsaures in that loop by more than 10K.3:6.F 4'.F.1.The reactor shall not bc operated vith one recirculation loop out of service for more than 24 hours.With the reactor operating, if onc recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITIOH vithin 24 hours unless the loop is sooner returned to service.1..Recirculation pump speeds shall bc checked and logged at least once per day.2.Folloving one pump operation, the discharge valve of the lov speed yump may not be opened unless the speed of thc faster pump is less than 50K of its rated speed 2.Ho additional surveillance required.3 When the reacti5mis not-4n the RUH mode, REhCTOR POWER OPERATIOH vith both recircu-lation pumps out-of-service for to 12 h'ing such interval, restart of 5g, the recirculation pumps ia g.], t.'f pcrmittcds provided the loop gq<c>aischirge temyerature is vithin 75 F of the saturation temperature of the reactor sg 3.0.R.'t 3.Before starting either recirculation pump during REACTOR POWER OPERATIOH/Al c s rgc tcmperaturc and dome saturation temperature. BPH Unit 2 3.6/4+6-12 AMENOMBlT g6.2 2 g PAGE~.OF Q s~)Sic.$(u~3.1 7 NR i 8 1993 3.6.F L~3.g,Q,f, gott 2 vessel water as determined by dome pressure.e a e apse t e in natural circulation and one pump operation must be no greater than 24 hours.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode.Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.J 5tc d~gk4j~~'c~ 40~c~r~Isi5 gc/, I 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of'ny ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a COLD SHUTDOWN CONDITION or less than 50'F above the minimum temperature required by NDT consider-ations, until each indication of a defect has been inves-tigated and evaluated. Inservice inspection of ASME Code Class 1, Class 2, and Class 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, Section 50.55a(g except where specific written relief has been granted by NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.'dditional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.II i'FN Unit 2 Stc.WHsgllcr,$ ~@g Cw5 gg g/qg Cg 3 6(4 6 13 AMENOMBF gy, p 06 PAGE~OF 0 ~rrarrrr okra%Sar~aaemIIaaaaaa ~aNIRIIaaarar aaaiislHrararr ~aaHIllirr rrrsllflrraraa aasSISaaaaaa ~rt)rlISrrraaa ~rQRRsaaaaaa ~allasaraaaaa ~aIIRSaaaaaaa E%Ãisrraaarr ~Ilasaaaaaaa 5I5!3~Ksa~~~..~s I I~~.~~~I~~~fl '4 W 3.6.B.1.PRIOR TO ST and at steami rates less than 00,000 Ib/hr, c folloving limits shall apply.4'.BE 1.Reactor coolant shall be c tinuously monitored or conductivity. except vhen there is no fuel in the reactor vessel.a.onductivity, pmho/cm at 25 C 2.0 Chloride, ppm 0 a.Whenever'he continuous conductivity monitor is inopcrablc, a sample o reactor coolant 11 be analyz for cond tivity every 4 rs except as 1 tcd belov.If the, cactor is in COLD SHUTDOWN COHDITIOEg a'ample of reactor coolant shall be analyzed for conductivity every 8 hours.b.Once a veek t e continuous nitor shall bc ected vith an in-li flov cell.This i ine condu ivity cal ration ahall be p formed every 24 ura vhenevei the reactor coolant conductivity is>1.0 pmho/cm at 25 C.2.At steaming rates greater than 100,000'b/hr, the folloving limits shall apply.a.Conductivity, idaho/cm at 25'C 1 0 During startup prior to pressurizing thc reactor above atmospheric pressurcg measurements of reactor vatcr polity shall be performed to ahov conformance vith 3.6.5.1 of limiting conditions. b.Chloride, ppm 0.2 BFS Unit 2 3.6/4.6-5 AMENpMENT RI.224 PAGF/OF~ P I'4~re rs'~war r I'I~v rr~~~-r<Tl? (~VF ttttl~a 1IhtrJttulLeLC)K~Llt%+I'0'k'6 I I~',.l'J:0~~~~~~~~I I~~~~~~I~I~I~~~~~~~~~'~I~~~~~~~II~~~~~~~'~~~~~~I~II I~~~~~~~~~'~~~~~~~~~~I~I I~~~~~~~~~~~~~~~~~~~~~~~I~~11 I~I I~~rr I~~I~II~~~~~~l'~I~~~~~~~~I~~l M~~~I~~I~~~~~~~~~~I I I~'I~~~~~~~~~I o

3.6.B.Cao an ist 4.6.B.Coa em st 4.When th reactor is not pr ssurized vith fuel in the r actor vessel, exce t dur the STARTUP CO TIOH, th reactor vater sha be intained vithin t olloving limits.a.Conductivit 10 pmho/at 25'C 4q enever the re ctor is not pressurized v h fuel i the reactor essel, a sample of e reactor coolant all be ana zed at lea every 96 ho rs for c cdcctivity, loride ion ontent and pH.b.Chlori-0.5 ppm c.pH s ll be betve 5.3 and Se6.0 5.When the time lim ts or um conduct ity or oride conc tration imits are ceded, an orderly shu ovn shall be initiated ediately.The reactor 11 bc brought to the COLD SHUTDOWH COHDITIOH as rapidly as.cooldovn rate permits.5.During equilibrium pover operation an isotopic analysis, including quantitative measurements for at least I-131, I-132, I-133, and I-134 shall bc performed monthly on a caolant liquid sample.6.ene e reactor is critical, thc limits an activity concentrations in the reactor coolant shall not exceed the equilibrium value of 3.2 pCi/gm of dose equivalent I-131.S<d'fdic~~<FiCV T<aM p'~A C RAhl g~~St h/lg Q~'~Y'4<s st=<7, g 6.Additional coolant samples shall be taken vhencver the rcactar activity exceeds onc percent of thc equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are.met: BFH Unit 2 3'.6/4'-1 AMENOMENT NL 2 24 pAGE 0 p if 0 <~$: 3.c.E/Mdiv JUN 2 8)PE 3.6.B.4.6.B.o 3.6.B.6 (Cont'd)This limit may bc exceeded folloving pover transients for a maxim'f 48 hours.During thfa activity transient thc iodine concentrations shall not exceed 26 pCi/gm vhcnever the reactor is critical.Thc reactor shall not bc opcratcd morc than SX of ita yearly povcr operation under this exception for thc equilibrium activity limits.If the iodine concentration in thc coolant exceeds 26 pCi/gm, the reactor shall be shut dovn, and the stem line isolation valves shall be closed immcdiatcly. X6C~TIE'ICWnoN wc~<~"~~R'4 814 Is~s 3.'/.6 i~~(g zpcvaQ 4.6.B.6 (Cont'd)a.During the STARTOP COHDITIOH b.Pollovtng a signfficant pover change**c.Folloving'n fncreaac in thc equflfbri~ off-gaa level exceeding 10,000 pCf/sec (at thc steam get air ejector)vithin a 48-hour period.d.Whenever the equilibrium iodine limit apecifi'cd in 3.6.B.6 is czcecded.Thc additional coolant liquid samples shall be taken at 4 hour intervals for 48 hours, or until a stable iodine concentration belov the limiting value (3.2 pCi/gm)is established. Hovcvcr, at least 3 consecutive simples shall be taken in all cases.kn isotopic analysis shall bc performed for each ample, and quantitative measurements made to determine the dose equivalent I-131 concentratio 7.Wh there is o fuel in thc r actor vesa , technic 1 pecificat n reactor oolant chemiatry fmits do ot apply.there is no fuel in thc eactor vcsa , aampli of reactor c ant ch try t technic specific ion frequency ia not rcquir d.**Por the purpose of this sectfon on sampling frequency, a significant povcr cxchu~e is defined as a change exceeding 15K of rated pover in less than 1 hour.BFH Unit 2 3.6/4.6-8 AMENOMEgl'ltd. 2 p g PAGE V OP l~ CTS R.C,g/+gg J NR i 8 Igga 3.6.F 3.6.F.3 (Cont'd)vessel water as determined by dome pressure.The total elapsed time in natural ci.rculation and onc pump operation must be no greater than 24 hours.Sce 34$4iPic4)jo~ peg W RFIV ISTIC 3.'f./Peci~c l 7 LooPS OpC~4~z i~/hit~ec4o~4.The reactoi shall not be operated with both recirculation pumps out-of-service while the reactor is ia the RUN mode.Followiag a trip of both recirculation pumps while in the RUN mode, imnediately initiate a manual reactor scram 3.6~G 4.6.G The structural 'egrity of ASME Code Clas 1, 2, and 3 equivalent components shall be maintai d in accordance with Spec'cation 4.6.G through t the life of the plant.a.ith the structural integrity of any ASME Code Class 1 equivale component, which is art of the primary sys not conforming to the above requirem s, restore the structural integrity of the affected omponent to within i.ts imit or maintain the react coolant system in either a COLD SHUTDOWN CONDIT N or less than 50'F above the minimum temperature req red by NDT consider-at'ons, until each indicatio a defect has been inves tigated and evaluated. Inservice inspection ASME Code Class 1, Class , and Class 3 components 11 be performed in acco ance with Section XI of t ASME Boi.ler and Pressure V esel Code and applicable A eada ae required by 10 CFR 5 , Section 50.55a(g except wh e specific written relief s been granted by NRC pure to 10 CFR 50, Section 50.55 (g)(6)(i). dditional inspections shall be performed on certain ci.rcumferent 1 pipe welds to prov e addi.tioaal prote ion against pipe w p, which could d gc auxili,ary control systems.BFN Unit 2 3.6/4.6-13 AMENDMBfT%7.2 0 6 pAGs+o~~

c~a.c.g/VC g.MAR I 8 1993 3.6.G 3.6.G.l ont'd)With the struc al integrity of any ASME de Class 2 or 3 equivalent omponent not conformi to the above requir ents, restore the struc al integrity of t'ff ted component to w in it limit or isolate t e a fected component fr m all 0 ERABLE systems.BFN Unit 2 3.6/O.6-l4 AMENOMENT NO.2 0 6 PAGE

During all mo es of operation, all snubber shall be OPERABLE except as oted in 3.6.8.1.All saf y-related snubbers are 1 ted in Plant Surv llance Instructions. 1 With one or more snubber(s) inoperabl on a system that is re ired to be OPERABLE in e current plant co ition, vithin 72 hours replace or restore th inoperable snubber(s) t OPERABIS status and erfora an engineer evaluation on the a tached component or decl re the attached system inoperable and follov the appropriate Limiting Condition statement for that system.Each safety-related snubber shall be demonstrated OHHtkBLE by performance of the folloving, augmented inservice insped'tion program and the requirements of.Specification 3.6.8/4.6.8. These snub rs are listed in Plant Su eillance Instructions. I ks used in this specification,"type of snubber" shall mean snubbers of the same design and aanufac er, irrespective of ca city.2~V Snubbers are ategorixed as inaccess le or acceseibl during reactor operatio.Each of these categor cs (inaccessible and ac essible)may be inspe ted independently according to the schedule determined by Table 4.6.8-1.The vi inspection interv for each type of ber shall be determined ed upon the criteria rovided in Table 4.6.8 and the first inspection nterval determine using this criteria shall be based upon th previous inspection interval as established by the requirements in effect'efore amendment No.225 BFH Unit 2 3.6/4.6-15 AMENOMEHT R0, p 25 PAGE~O~~ 4.6.H.3~0 BPS Unit 2 3.6/4.6-16 Vis inspections shall v fy-that (1)the abber no visible indi tions of dsaae or iepai ed OPERLBILITY; (2)attachments to c foundation o supporting structure e functional, and (3)teners for the etta'cha t of the snubber to th component and to the snub r anchorage are f tional.Snubbers vhich a ar inoperable as a esult of visual inspectioni shall b classified unacce able and aalu be reclassif ed acceptable for e purpose.of establis the next visual insp ction interva.provided t (1)the cause of the r/ection is clearly establ hed and reaedied for t particular snubber and or other snubbers irr pective of type t be generically susceptible; and)the affected snubbe is functionally sted in the as-found co ition and detezained PERABLE per Specific ion 4.6.8.5.k review evaluation shall be per ormed and documented to)tify continued ope tion with an unacceptable snubber If continued operatio cannot be/ustified, th snubber shall be decla d inoperable the LIMITIEG COSDITIOHS R OPBIRTIOH shall be me OMENS NJ;p 25 pAGE~OF~ 0 4.6.8.3 (Cont'd)~k tionally, snubbers tached to scctio of safety-related s tees that have experienc unexpected potentially aaaglag transient since the last inspect n period shall bc evalua ed for the poss ility of concealed d e and functionally tc tcd, if applicable, to nfirm OPBRABILITY nubbcrs which ha been made inoperable thc result of un ected transients, solated daaage, o other random events, en thc provisions of 4.6.7 and 4.6.H.have been ct and any o r app opriate corre ive ac ion implemen d, shall not be counte in determining c next visual inspection terval.BFH Unit 2 3.6/4.6<<17 NENOMENT Ng.p g 5 PAGEOS~

4.6.H.~S~~s~4~ing each refuel tagc, a reprcs ative sample of 10K o the total of each type o safety-relet d snubbers in use in thc lant shall be functio ly tested either in pla or in a bench test.The representative sample a ected for functional eating shall include the various configuratio operating environm ts, and the range of size and capacf ty of scc srs sithin the types.rcprescntat e sample should be eighed to.include orc snubbers from severe crvicc areas such as n r heavy equipment. stroke setting d thc security of fast rs for attachment of th snubbcrs to thc compon and to the snubber anch age shall be verified o snubbcrs selected or FUKCTIOKlL TESTS.BFH Unit 2 3.6/4.6-18 AMENDMENT go.p p 6 Gp~dOP~~ 0 JAN 18 1888.6.H.5.e snubber FUHCTIOHhL TEST shall verify that: 4 a.hctivation (re raining action)is a ieved in both te ion and compress n vithin the specif ed range, except tha nertia dependent, a eleration limiting echanical snubbers may be tested to verify only that activation takes place in both directions of trav b.Snubber bleed, r release vher required, is present n both compress n and tension vithin he specified rang o c.or mechanical snubbers, the force required to initiate or maintain motion of the snubber not great enough to overstress the a ached piping or comp ent during, the movement, or to ind ate impending failure f the snubber.d.Fo snubbers specifically quired no t to displace under continuous load, the ability of the snubber~to vithstand load vithout displacement shall be verified.BFN Unit 2 3.6/4.6-19 AMENDMENT NL l 6 9 PAG~~r't:~8' C C7S S.g.g 6.g P JUL 055%.6 H.4.6.8.5 (Cont'.eating methods may be used to measure parameters indirectly or parameters other than those specified if those results c be correlated to the specified param ers through estab shed methods.6.ka ineering evaluation 1 be made of each failure meet the PORCTIOEhL TEST acceptance criteria to determine the cause of e failure.The result this analysis shall be, if applicable, in sel cting snubbers to be t ted in the subsequent lot n an effoxt to determine th OPERABILITY of other snub rs which may be sub)ect the same failure mode.election of snubbers for f ure testing may also be bas on the failure sis.For each snubber t does not meet the CTIOKAL TEST acceptance criteria, an additional lot equal to 10 percent of remainder of that type f snubbers ahall be fun ionally tested.Testing s 1 continue until no dditional inoperable snubb s are found vithin subsequ t lots or all snubbers of e original tOICZZOELL tyya hare been)tested or all suspect snubbers identif d by the failure analys s have been tested, as applicable. BFH Unit 2 3.6/4.6-20 ~AIH<n~an e.p pq PAGE~iOF/ 4.6.H.4.6.H.6 (C'd)If any snubber selected for functional test either fails to croup or fails to e, i.e., frozen in ace, the cause vf.ll b valuated and if caus by manufacturer or d ign deficiency, all snubbers of the same design subject to the same defect shall be functionally tested.This testing requirem shall be independen f the requirement tated above for s ers not meeting e FUNCTIONAL TEST cceptance cri,teria. The discovery of loose or missing attachment fasteners vill be evaluated to determine vhether the cause may be.localize or generic.The res of the evaluation'l be used to sel other suspect ubbers for veri ng the attachment teners, as applicable. 7.For the snubber(s) und inoperable, an e neering evaluation s be performed on the co nents vhich are restra ed by the snubber(s). Th urpose of this ineering evaluation shall be to determine if the components restrained by the BFN Unit 2 3.6/4.6-21 NmoMen R.I SO waco~~ JAN f 9 1888 4.6.H.4.6.H.7 t')snubber(s) e adversely affecte y the inoperability of snubber(s), and in orde ensure that the restrained component remains capable of meeting the designed servic 8~u b s Snubbers v fail the visual inspect or the FUN NAL TEST acceptance c teria shall be repaired or replaced.Replacement snubbers and snubbers Which have repairs vhich might affect the FUNCTIONAL T results shall meet t FUNCTIONAL TEST teria before instal ion in th>>unit.Th snubbers shall have m the acceptance cr ria subsequent to their st recent service, and the FUNCTIONAL TEST must have been performed vithin 12 months before being install in the unit.9.V ua anent or other exemptions from visual inspections and/or functional testing for individual snubbers y be granted by the C ssion if a justifiable asis for exemption is esented and if appli e snubber life dest ive:esting v performed to qualify snubber OPERABILITY for the applicable design conditions at either the BFH Unit 2 3.6/4.6-22 @t)MENT NO.X 6 0 PAGp~o~

4..H.4.6.8.9 (Cont'pletion of their fabrication or at a s sequent date.Snubbers so empted shall continue be listed in the plant tructions with footnotes icating the extent o the exemptions. e service life of sn bees may be extended base on an evaluation of the ecords of FUNCTIONAL TEST maintenance h tory, and environment conditions to vhich the ubbers have been exp sed.BFN Unit 2 3.6/4.6-23 AMENOh)BIT ND.X 6 0 PAGp~50F~

'JAN 19$888 THIS PAGE IHTEHTIORALLY LEFT BLANK BEE Unit 2 3'/4.6-23a NENtjMENT NO.X 6 0 r s a~~koF~S ble 4.6 H-1 SHUBBER SUAL IHSPECTIOH IHTERVAL Population or Catego Notes Column A Extend Interval Column B Repeat Int al 4 Column C Reduce Interval e Hote 1: The next visual inspection interval or a snubber population or category size shall be determined ased upon the previous inspection interval and the n er of unacceptable snubbers found during that interval.ubbers may be categorized, based upon their accessibility d ing pover operation, as accessible or inaccessible. These tegories may be examined separately or)ointly.Hovever, e licensee must make and document t decision before any pection and shall use that decisio as the basis upon wh to determine the next inspection erval for that categ Hote 2: Interpola on between population or category siz and the number unacceptable anubbers is permissible Use next lover integ for the value of the limit for Col A, B, or C if tha integer includes a fractional value unacceptable snu ers as determined by interpolatio Hote 3: If the number of unacceptable snub rs is equal to or less than ,the number in Column A, the nex inspection interval may be twice the previous interval not greater than 48 months.BFH Unit 2 3.6/4.6-23b FINED'8~(

T le 4.6.8-1 (Continued) S BBR VISUAL IHSPBCTIOH IHTERVAL Hote 4: If th umber of unacceptable snub rs is equal to or less the number in Column B but eater than the number in Co A, the next inspection terval shall be the same as e previous interval.Hote 5 If the number of unacce able snubbers is equal to or greater than the number in Co C, the next inspection int al shall be two-third of the previous interval.How er,'if the number of unacce able snubbers is less than the umber in Column C, but eater than the nuaber in Col , the next interval be reduced proportionally by terpolation, that is, previous interval shall be r ced by a factor that is e-third of the ratio of the d ference betveen e number unacceptable snubbers found uring the previo interv 1 and the number in Column B o the difference the numbe in Columns B aud C.Hote 6: The provisions of Specificatio 1.0.LL are applic le for all inspection intervals up to including 48 mon BPH Unit 2 3.6/4.6-23c AMENOMENT RQ.p Zg p.<gE~$'F~3' UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF g

'n LIMITIHt COHDITIOHS FOR OPERATIOH SURVEILLAHCE REQUIREMEHTS

4.6 Applies

t the operating s atus of the res or coolant syst plies to the p iodic ination and t ting req rements for th reactor cool t system.To assure the integr and safe operation of the react r coolant systeme To dete the condition of the reactor oolant system d the operation f the safety devices relate to it.1.The average rate of DO S't.g reactor coolant temperature change during normal heatup t or cooldoma shall not exceed 100 F/hr shen averaged over a one-hour period.ill!<Mgcooldoms, the c ggAQ,)following parameters shall be recorded and reactor coolant temperature determined 36 ute intervals til 3 succ s ve readings a each given lo tion vithin F.a.team D e Press e (Conve to upp vess region t erature)b.actor ttom drain empera re Reci culatio loops A B d.eactor v ssel bottom head tern erature e.React vessel shell ad)a ent to shell fl e BFS Unit 3 3.6/4.6-1 PAGE~pp~ 3.Daring heatap by nonnuclear means, exccyt when the vessel is vented or as indicated in 3.6.h.4, daring cooldown folloving nuclear shutdown, or daring lov-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve$2 of Figure 3.6-1 antil removing tension on the head stad bolts as specified in 3.6.L.5.LCo 3A g 2.During all operations vith a critical core, other than for lov-level pbysics tests, LCg.except vhcn thc vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temycratare of carve 03 of Figure 3.6-1.5 3'.M..eactor vcsscl metal temperature at the outside surface of the bottom head in thc vicinity of thc control rod drive housing and reactor vessel shell ad)accnt to shell flange shall be recorded at least every LZ minutes daring ervice hydrostatic or test~cn the vesee 5g g,q,fj,)~ressurc is>312 psig 3.Test specimens reyresent the reactor vessel, b e veld, and veld hea affected zone metal I be ins alled in th reactor v sel aQ ent to the cssel at the c c mi plane level.The be and type of sp im vill be in acc vith GE reyort lKDO-10 15.The specimens shall meet the intent of hSTN E 185-82.BFH" Unit 3 3'/4'-2 PAGE OP~AMENDMENT NO.14 1 l.i <Pc ca'P i 3 1995 5$'3A.g,)NnH'.g, 4.The beltline region of reactor vessel temperatures LC.O (during inservice hydrostatic 9'g.9 or leak testing ahall be at or above the temperatures shown on curve¹1 of Figure 3.6-1.The applicability of this curve CD Chese.Ccats is cztended to ncmnuclear heatup and aabient loss cooldom associated with these tests only if the heatup and cooldem rates do not exceed 15 F per hour>>SRg.g.g.l Nog)p)y p~g sC.9.'L9.2.~R~~9~9e (>>+bloke, sg z.q,9.a+.Mom LC.b 3 9.1 5.The reactor vessel head bolting studs may be partially tensioned (four sequences of the seating pass)provided Che sCuds and flange.IaCerials are above 70 F.Before loading the flanges any more, the vessel flange and head flange must be greater Chan 70 F, and Itust reaain above 70 P whfle under fuX1 tensicm.5.When the reactor vessel head bolting studs are tensioned snd the reactor is in a Cold Condition, the reactor vessel t eratur imediately below'he head flange shall be Prod~w~c~Pr 5',g,),5>PAGE OF BFK Unit 3 3.6/4'-3 AMENDMENT Ng.y 9 8

ecifim'3 I~+@i 6.The pump in an idle recirculation loop shall not be started unless thc temperatures of thc coolant vi in thc idle o erat ng recirculation loo are w 5D F of each other.+3'.o9.6.PJo)r urfag ST of an idle rec rculation loop, the temperature of the reactor coolant the operat and die loops l bc+y o c I Co$,4,cj 7.The reactor recirculation pumps shall not be started unless the coolant temperatures betveen the dome and bottom head drain are vithin 145 F.Prior to atartiag a recircu ation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall bc compare M3 p~~>~Pcy IeN$p, g~~PAGE OP~BPS Unit 3 3.6/4.6W

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4.6.E.~Je~~I Sc<'Sus+t Ei'rc4onQ~ c~ey+~>F'N iSTS 3q.~I 5<@5gs+iCi ca/ion Q~cha<g~<kr'P'hl l5T5 SI I 2.Whenever there is recirculation flov wit the reactor in the STARTUP or RUN Mode and one recirculation pump is operating, the diffuser to lover plenum differential pressure shall be, chcckcd daily and the differential pressure of an individual jet pump in a loop shall not vary from thc mean of all jet pump differential pressures in that loop by more than 10'.3.6.F at 0 4.6.F e 0 tio 1.The reactor shall not bc operated vith onc recirculation loop out of service for more than 24 hours.Pith the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWH COHDITXOH vithin 24 hours unless thc loop is sooner returned to service.1.Recirculation pump speeds shall be checked and logged at least once per day.sR9.g 2.Folloving onc-pump operation, the discharge valve of the lov speed pump may not be opened unless thc spccd of the faster pump is less than 50K of its rated speed.3.When thc reactor is not in thc RUH mode, REhCTOR POWER OPERhTIOH with both recirculation pumps out-of-service for up to 12 hours i ermitted Dur ng such interval restart of thc recirculation pumps is permitted, provided the loop discharge temperature is vithin 75 F of the saturation temperature 2.Ho additional surveillance required.~R 3.Before starting either recirculation pump during REhCTOR POWER OPE&Bd'e loop dis arge temperature and dome saturation temperature. pAGE~OF~BFH Unit 3 3.6/4.6-12 AMENDMgfT NP.y 8$ .6.F 3,9A,9~Ivo&2.of the reactor vessel water as determined by dome pressure.The a e apse t e n natural circulation and one pump operation must be no greater than 24 hours.See V<S+Pi~gon kr C$r Bpe Isis r,v.j 4.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor's in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.3.6.G 4.6.G The structural integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained in accordance with Specification 4.6.G throughout the life of the plant.a.With the structural integrity of any ASME Code Class 1 equivalent component, which is part of the primary system, not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or maintain the reactor coolant system in either a Cold Shutdown condition or less than 50 F above the mintunun temperature required by HDT consider-ations, until each indication of a defect has been investigated and evaluated. 1.Znservice inspection of ASME Code Class 1, Class 2, and~Class 3 components shall be performed in accordance with Section XX of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by HRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). 2.Additional inspections shall be performed on certain circumferential pipe welds to provide additional protection against pipe whip, which could damage auxiliary and control systems.3ietifiaaiion 6r Cjgunyg for C T5 y,i, p./q BFH Unit 3 3.6/4.6-13 AMENOIHEgf gg, y p g A8~$OF~~ .8 V.S.Z JUN 2 8 199$3.6.B.1.PRIOR TO ARTUP and at steam rates less t 100,000 lb/hr the following limi s shall apply.4.6.BE l.eactor coolant shall be continuously monitored for conductivity except when there is no fuel in the reactor vessel.a Conductivity, pmho/cm at 25 C 2 b.Chloride, ppm 0.1 I~I a.Whenever ghe continuous onductivity monitor inoperable, a samp of reactor cool shall be ana zed for c ductivity every hours except as listed below.If the reactor is in COLD SHUTDOWH COHDIT105, a sample of reactor coolant shall b analyzed for conductivi every 8 hours.b.e a week the ontinuous monitor shall be checked with an in-line flow cell.This in-line conductivity calibration shal be performed eve 24 hours whenev the reactor coo t conductiv is>1.0 ymho/cm t 25iC, 2.kt steaming rate greater than 10 ,000 lb/hr, the fo owing limits shal apply.a.Conductivity, pmho/cm at 25'C 1.0 2.During st tup prior to pressur ing the reactor'bove tmospheric pre ure measurements o reactor water quality hall be performed to show conformance with 3.6.B.1 of limiting conditions. b.Chloride, ppm 0.2 BPK Unit 3 3.6/4.6-5 AMENDS@~0.I 8~AGE/OF Ci CTS Z.(,8 q.a.g OEC 0 7 1994 4.6.B.3~ht steaming rates greater than 100, 0 lb/hr, the react water quality exceed Specifi tion 3.6.B.2 only'r the time limits pecified belov.Exc eding these time limi or the folloving maximum ality limits shall be caus for placing the re tor in the COLD OWH CORD IOH.Conductivity time above 1 idaho/at 25iC-2 ve/year.Limit'0 o/cm at 25 C 3~Whenever the reactor is operating (including HOT SThHDBY COHDITIOH) measurements f reactor vater quali shall be performed ccording to the foll ing schedule: a.oride ion content d pH shall be measured at least once every 96 hours.b.Chloride ion content sha be measured a least every 8 h rs vhenever reactor conduc vity is>1.0 o/cm at'C.b.oride concentration ti e above 0.2 p 2 vecks/ye Maximum Lim Oo5 ppme c.sample of reactor coolant shall be measured for pH at least once every 8 hours vhenever the reactor coolant conductivity is>1.0 idaho/cm at 25iC.c.The rea or ahall be placed in the SHUTDOWNS COHDI IOH if pH<5.6 or>8.6 for a 24-hour period.BPH Unit 3 3.6/4.6-6 N>NEON.I 86 PAGE

3.6.B..6.B.Coo t st 4.When the eactor is not pre urized vith f 1 in the re tor vessel, cept duri the SThRTUP HDITIOH, the eactor vater 11 be mai tained vithin the fo loving limit.Conductiv ty-10 pmho cm at 25 4.Whenever t e react r is not pressuri d vith uel in the rea or ves 1, a sampl of the eactor cool t shall e analyzed at east ev 96 hours f conduc vity, chloride on conten and pH.b.Chio de-0.5 c.p shall be tveen.3 and 8.6 5.Wh the tim limits or co ctivity or oride c ncentration limits ar exceeded, an ordirly hutdovn shall be initia ed immediately. The react r shall be brought to the OLD SHUTDOWH COHDITIOH as apidly as cooldom rate permits.5.During equilibrium pover operation an isotopic analysis, including quantitative measurements for at least I-131, I-132, I-133, and I-134 shall be performed monthly on a coolant liquid sample.6.enever e reactor is critical, the limits on activity concentrations in the reactor coolant shall not exceed the equilibri~ value of 3.2 pCi/gm of dose equivalent I-131.S<c 3'u.~f;~/on QP.C~ge<t S~nt isis s,g,g Section 6.hdditional coolant samples shall be taken vhenever the reactor activity.exceeds one percent of the equilibrium concentration specified in 3.6.B.6 and one of the folloving conditions are met: BFH Unit 3 3.6/4.6>>7 AtaENoMENT go.y 8 y PAGE

3.6.B.3.6.B.6 (Cont'd)This liait Nay be exceeded folloving, pover transients for a aaxiarm of 48 hours.During this activity transient the iodine concentrations shall not exceed 26 pCi/gw vhenever the reactor is critical.The reactor shall not be operated aore than 5Z of its yearly pover operation undir this exception for the equilibrium activity liaits.If the iodine concentration in the coolant exceeds 26 pCi/ga, the reactor shall be ahut dovn, and the steaa line isolation valves shall be closed iwaediately. See V~q6<i'm~n $r.~~~n>s Vs r.q,(.;~;, Scc ho g 4.6.Bi 4.6.B.6 (Cont'd)a.During the STARTUP CORDITIOK b.Folloving a significant pover chLnge**c.Follovtng.an increase in the squilibritm off-gas Xerel exceeding 10,000 pCi/sec (at the steaa)et air e)ector)vithin a 48-hour period.d.Whenever the equilibrhm iodine liait specified in 3.6.B.6 is exceeded.The additional coolant liquid saayles shall be talon at 4 hour'ntervals for 48 hours, or until a stable iodine concentration belov the liaiting value (3.2 yCi/ga)is established. Hovever, at least 3 consecutii saaples shall be taken in all cases.An isotopic analysis ahall be perfozaed for each sawple, snd quantitatire aeasureaents aade to deteraine the dose equiralent I-131 concentration. 7~the e is fuel in the actor esse , te cal pecif cati reac r coolant cheai try 1 ts do not apply.7.en ther is fuel e reac r v sl)saapl of r ctor c olant cheai ry a tschni 1 spec ficat on fra ency is t requi ed.For the purpose of this section on saapling frequency, a significant pover exchange is defined as a change exceeding 15Z of rated pover in less than 1 hour.BFN Unit 3 3.6/4.6-8 NENBMgpgQ, z gz'AGE~OF~~ tl .6.F 0 a 3',F,3 (Cont'd)of the reactor vessel water as determined by dome pressure.The total elapsed time in natural circulation and one pump operation must be no greater than 24 hours.see swiJ"cqAo* k, C+~~B~H>~T~3'~~Pgcirculq~ 9 p s a+His 5ecti on 4.The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUH mode.Following a trip of both recirculation pumps while in the RUH mode, immediately initiate a manual reactor scram.~6.G S U 4.6.G The struc ral integrity of ASME Code Cla 1, 2, and 3 quivalent compon s shall be intained in acc dance with S ecification 4.6.G hroughout t life of th plan~a.With the s ctural int rity of any Code Class equival t component, which is pa of the pr system, not onforming to e above re irements, res ore the s ctural inte ity of the ffected compo ent to with its limit or intain the reactor coo t system i either a Cold Shu own conditi or less t 50'F abov the min temperatu e~requir by HDT cons der-atio , until each ndication of a defect has be n investigated and evaluated. 1.nservice pection f ASME Code Clas 1, Class , and Class 3 omponents hall be perform d in accor ance with Sectio XI of the ASME Boiler and P essure Ves el Code d appl cable Add a as re ired by CFR 50, ction 50 55a(g)ex ept where ecific itten r ief has b grant by HRC rsuant to 0 CFR 50, Section 0.55 (g)(6 (i).2.Additio inspect ons shall be perform on cert in circum rential ipe welds to pr ide addi ional prot tion aga nst pipe whip, whi could d ge auxiliary and control s stems.BFH Unit 3 3.6/4.6-13 AMENDafEHT NO.r 7'9 PAGE~OF~~ a vs 3.6.g./q, 0.4-NY31m 3.6.G 3.6.G.l (ont'd)4~6.G b With t structu al integ ity of ASME Co Class 2 or 3 equ alent co onent no co forming t the abov r quirement , restor the tructura integrity of the affected omponent o vith its lim or isol e the affect d compon from l OPE syst BFH Unit 3 3+6/4+6-l~ AMENDMENT NQ.g 7 g PAGE~OF/ 0 3.6.8.4QQhhCXk 4 6 8 2m8hma During all modes operation, all saubbcrs sha be OPERABLE except as noted n 3.6.H.1.hll safety-rcl ed snubbers are listed in lant Surveillance astructions. 1.With o or more snubb (s)inoperable a a sys em that is re red to b OPERhSLE in cur cnt plant cond ion, wi in 72 hours r place o restore the i operablc ubber(s)to 0 RABLE~tatus and per orm aa engineeriag c aluation on the atta cd component or declare e attached system ino erable and f olla appropriate Limiting Condition statement for that system.Each sa ety-rclatcd snubber shall e demonstrated OPE by performance of e folloving augmented i rvice inspection program thc requirement of pecification 3.6.4.6.8.These snubbers ar listed ia Plant Surveill e Instructions ka use in this speci icatioa,"type of snub er" shall mean sn bcrs of the same d ign and manufac rcr, respective of c acityo 2~Snubbcrs are ategorizcd as inacccss le or accessible uring reactor operation Each of these categori s (inaccessible and ac ssible)may inspe ed indepcnd tly acco ing to the s cdulc det rained by Tab e 4..8-1.The v ual pection int al for each type of ubber shall be determin based upon the criteri provided in Table 4:6.-1 aad the firs iaspecti interval determi ed using this criter a shall be based upon c previous insp ction interval as est lishcd by the req ircmcnts in cffcct before amendmcat No.183 BFH Unit 3 3.6/4.6-15 IIILENOMENT No, y 88 PAGE~OF

Visual inspectio shall verify that (1)e snubber has no visibl indications of Cease o layaired OPERABILI~(2)attachR s to the founda on or supporting struc re are functional, and 3)fasteners for the at chment of the snubber t the component and to the ubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified unaccep able and aay be reclassif d acceptable for e purpose of establish the next visual insp ticm interva.provided t (1)the cause of the r ection is clearly establi ed and reaedied for t particular snubber and or other snubbers ir spective of type that be generically susceptible; and (2)the affected snubber is functionally tested the as-found conditicm detezained 0 per Specification 4.8.5.reviev and ev uation shall be perforae snd docuaented to gustif continued operatio with an unacce able'snubber. If ccmt ued operation cannot be Justified, the snubber shall be declared inoperable and the LINZTIEC COHDITIOKS FOR OPERLTIOK shall be aet.BP5 Unit 3 3.6/4.6-16 AMENDMENr ND.Z 8g PAGE 9 pp/g 4 6 8 Bmhhern 4.6.8.3 (Cont'd)kd tionall , snubb rs a tached sectio of afety-r ated s teaa that have erience unexpected poten illy d iag tr eats s ce the ast ins ection riod 11 be ev lusted or the ssibili y of c cealed e func onally tested if app cable, to conf i OPB LITT.Snab ers vhi have be sad inoper le as th re t of expected t ansien , isolat aaage, r other oa events when the rovisions of 4..8.7 and.6.8.8 ve be met and other ap opriate rrectiv a ion 1ap1 ented, 1 ot be co ted in deterain the n t visual inspect interv 1.a BFH Unit 3 3'/4.6>>17 AMENOMBfT NO.I 8 3 PAGE~oF~S

4.6.8.4.Daring ch refue ng outage a repres tative aaapl of lOX the total of e ch type aa ty-relat snubbera in e in the ant shall be unctional teated either in place r in a ben teat.The re eaentativ aLRple aele ed for fun ional tes ng shall 1 elude the va ious confi rations, o crating en ironILenta, and e range o size and capacity snubbers ithin'he type.The repres tative s pie should e veigh d to inclu e aore ubbers frc seve e aervi e areas such aa ear he equfpaent. The atro e setting the aecuri of fastener for atta ent of the bbera to e coaiponent to the anu ber anchorag shall be ve fied on anu era selected for CTIOKAL TESTS'FR Qnit 3 3.6/4~6-18 em0mHrHU X83 PAGc~oF-4.6.H.S u e 5.snubber CTIOHhL ST shall v rify that: a.Activ ion (restr ning=acti)is achie d in oth tension and co pression vi in the ecified r e, except hat inerti dependent, accelerati limiting mechanic snubbers may be test to verify only t at activation takes place in bot dir tions of trav l.b.S ubber bleed, r elease where equired, is present i both compression and tension within the specified range+c.For me ical snu ers, the rce require to ini ate or mai ain mo ion of the ubber is t great eno to verstress e attached piping or omponent during t rmal movem t, or to i icate imp ing failur of the snu er.d.For nubbers sp ifically re ired not t displace er continu load, he ability f the snubbe to vithst load vithout displacem t shall be verified BFH Unit 3 3.6/4.6-19 NENDMENT NO.1 3 4 e Ga~l'r~8' 0' 4.6.8.4.6.H.S (Co?Lt'd e Testing methods may be used to mLcasurc p raactcrs indirectly or aactera other than sc spccif ic if those suits can be correl ed to the aye iad parameters ough established aethoda.6 X~a&a BFS Unit 3 tel 3 6/4.6-20 ka engineeri evaluation shall be de of each failure to ecc e PUECTIOKAL TEST acce ance criteria to d I%inc the cause of the ailure.The result of this analysis shall be used if applicable, in aele snubbera to be t ted in the subsequent lo an effort to deterainc OPERhBILIIT of other an era which say be sub/a to the aaae failure Selection of snubbera r future testing aalu also be aaed on the failure analysis.For each ubber that does not aee the lUKCTIOML YES acceytance criteria, ddltional lot equal to yercent of the reaa r of that type of snub ra shall bc functi ly tc ed Tcsti?Lg s ontinue until no itional inoperable anu era are found within suba cnt lots or all snubbcrs the original tUKCTZ TEST type have bean i teat or all suspect ambbers identified by the failure analysis have been teated, aa apylic able.NENOMENT N.y 83 pAG<~Ocean l8'

19 1989~6 oHo'gt~4.6.H.6 (C d)If any snubber ected for functi testing either ils to lockup or ls to move, i.e., ozen in place, the caus will be evaluated and if caused by manufac er or design defici y, all snubbers the same desig bject to the s defect shall be ctionally tested.This testing requirement shall be independ t of the requireme stated above for bbers not meetin e FUNCTIONAL TES cceptance criteria.e discovery of loose or missing attachment fasteners will be aluate to determine w er the cause may b ocalized or generic e result of the luation will be u to select other suspect snubbers verifying the achment fasteners, applicable. 7~ed 1 b nts r the snubber(s ound inoperable, gineering evaluation all be perform on the c ponents which are restr ned by the snubber(s The urpose of this e ineering evaluat shal o determine if e compone restrained b he BFN Unit 3 3.6/4.6-21 N ENDMQPNP Zeg PAGE~~~P~ cTS 3.4.'JAN 19 1"-P.6.H.~~~s 4.6.8.7 (Co d)snubber(s) vere ersely affected by e inoperability of the s ber(s), and in order to re that the restrained co onent remains capable of ecting the designed ice.8.S ubb s Snubbe vhich fail the visual insp tion or the CTIONAL TEST accepta e riteria shall be re red or replaced.Rep cement snubbers and s hers vhich have repair hich might affect t FUNCTIONAL TEST result shall meet the FUN OHAL TEST criteria b ore installation i he it.These snubb s shall have met the ac ptance'riteria sub quent to their most rec service, and the FUNCTIO TEST must have been rformed vithin 12 mo before being ins led the unit.C 0~J BPN Unit 3 3.6/4.6-22 Pe ent or other exemptions fr visual inspections d/or functional testing for individual snubbers y be.granted by the C ission if a Justifiable asis for exemption is esented and if applic e snubber life destruc ve testing vas rformed to qualify s ber OPERABILITY for e applicable design conditions at either the NENDMENr NK X ac pAgE OF CVg Z.6 JAR i 9 1988 4.6.H.~t~4.6.H.9 (C'd)completion of thei fabrication or a subsequent date.Snu rs so exempted shall c inue to be listed in t plant instructions with f tnotes indicating th extent of the exemp ns.10.S e ora The se ce life of snubbers may e extended based on aluation of the rec s of FUNCTIONAL TESTS, maintenance his ry, and environment conditions to vhich the nubbers have been osed.BFN Unit 3 3.6/4.6-23 AMENOMENT NQ.X3 pwaa~~o>~~: Cl 0 mrs.c.a/v<.e JAN I 9 1989 THIS PAGE IHTEHTIOHALLX LEFT BLAHK BFH Unit 3 3.6/4.6-23a ANENOMENT Ng.X'3PAGE~6'F~~ Table 4.6.8-1 SHUBBER VISUAL IHSPECTIOH IHTERVAL Populati or Cate ry 0 Col A Extend terval Column B Repeat Interval Coluan C Reduce Interval Hote 1: The next visual ection interval fo a snubber population category size sha be deterained bas upon the previous inspection inte al and the amber unacceptable snubbers found during t interval.Snubb s aalu be categorized, ased upon their ac essibility during er operation, as aces ible or inaccess le.These categori s aay be exaained sepa tely or/ointly Heaver, the lic ee auat aake and docua t that decision efore auy inspecti and shall use that de sion as the baai upon which to dete e the next inspecti interval for category.Hote 2: Inte lation between po ation or category siz s and the n er of unacceptable ubbera is peraissible. Use next'lower in cger for the value f the liait for Col A, B, or C if t integer includes a fractional value of cceptable ubbers as detera d by interpolation. Hote 3: If the nuaber of cceptable snubbers is equal to or lese than the nuaber in Co A, the next inspect on interval cay be tvice the previ interval but not gre ter than 48 aonths.BPH Unit 3 3 6/4.6-23b AMENDMENT N, g 83 Table 4.6.8-1 (Continued) SSUBBER VISU ISSPECTI05 I hL If e nuaber of cceptable snub rs is equal o or less the number in Colum B but g eater than nenber i Co umn k, the nex inspection in erval shall b the saa as e previous int rval.f the nuaber f unacceptabl anubbera ia e to o.greater than the n~r in Column C the next inap ction in rval shall be tv thirds of the revioua inte al.Ho@er, if the nuaber of cceytable bere ia leaa the er in Column C, t greater the member Golda, the next interval 1 be reduc proporti by int rpolation, that is, the previous terval ahall e reduc by a facto that is one-third of e ratio of e differ ce between e nuaber of unaccepta e anubbera fo dur the previo inte al and the n er in Column to the ifference the n~rs in Col B and C.yrovfakoma f Specificati 1 O,LI re applicab for all inspection int rvala up to including 48 aontha.O'I 3 6I4.6-23c NENOMEgr gp.y 88

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEMPERATURE (P/T)LIMITS MINIST TIVE Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore,'nderstandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. ~A~Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.These surveillances are a duplication of the regulations found in 10 CFR 50 Appendix H.These regulations require licensee compliance and can not be revised by the licensee.Therefore, these details of the regulations within the Technical Specifications are repetitious and unnecessary. Furthermore, approved exemptions to the regulations, and exceptions presented within the regulations themselves, are also details which are adequately presented without repeating the details within the Technical Specifications. Therefore, retaining the requirement to meet'he requirements of 10 CFR 50 Appendix H, as modified by approved exemptions, and eliminating the Technical Specification details that are also found in Appendix H, is considered a presentation preference which is administrative in nature.A3 For clarity, the terms"prior to and during startup" and"prior to" have been replaced with"15 minutes".This Frequency is effectively the same since the proposed Surveillance now must be performed no more than 15 minutes prior to startup of the idle recirculation loop.This is essentially equivalent to the current requirements. I BFN-UNITS 1, 2, 8E 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEHPERATURE (P/T)LIHITS A4 e Proposed SR 3.4.9.4 requires verification that the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature are within 50 F of each other.CTS 3.6.A.6/4.6.A.6 requires verification that the temperatures between the idle and operating recirculation loops are within 50 F of each'ther. The temperature of the"operating recirculation loop" is considered equivalent to the RPV temperature. Therefore, this change is considered administrative. Proposed SRs 3.4.9.5, 6 5, 7 require the reactor vessel flange and head flange temperatures be verified>82 F, while CTS 4.6.A.5 requires the reactor vessel shell temperature immediately below the head flange be recorded.The BFN procedure that implements this requirement requires the vessel flange and head flange temperature be verified and requires the shell temperature be recorded.Since the intent of the surveillance is to verify vessel flange and head flange temperature to satisfy CTS 3.6.A.5 and both the current and the proposed SRs do this, the two are considered equivalent. As such, the proposed change is administrative. TECHNICAL CHANGE-NORE RESTRICTIVE A new Surveillance Requirement has been added.SR 3.4.9.2 ensures the RCS pressure and temperature are within the criticality limits once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality. This is an additional restriction on plant operation. H2 Three new Surveillance Requirements have been added.SR 3.4.9.5 ensures the vessel head is not tensioned at too low a temperature every 30 minutes.SRs 3.4.9.6 and 3.4.9.7 ensure the vessel and head flange temperatures do not exceed the minimum allowed temperature. These are additional restrictions on plant operation since the current requirements have no times specified. M3 ACTIONS have been added (proposed ACTIONS A, B, and C)to provide direction when the LCO is not met.Currently, no real ACTIONS are provided.These ACTIONS are consistent with the BWR Standard Technical Specification, NUREG 1433, and are additional restrictions on plant operation. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE~OF~

JUSTIFICATION FOR CHANGES BFN ISTS 3.4.9-RCS PRESSURE AND TEMPERATURE (P/T)LIMITS TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Details of the methods for performing Surveillances, and any, requirement to record data, are relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. Changes to the procedures will be controlled by the licensee controlled programs."Specific" The Frequency of this Surveillance has been changed from 15 minutes to 30 minutes.Verification that RCS temperature is within limits every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes is reasonable in view of the control room indication available to monitor RCS status.Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. In addition, this new Frequency is consistent with the BWR Standard Technical S ecification, NUREG 1433.Ll P'I L2 The Frequency of this Surveillance has been changed from 15 minutes to 30 minutes.The metal temperature is not expected to change rapidly due to its large mass, thus a 30 minute Frequency is adequate.In addition, this new Frequency is consistent with the BWR Standard Technical Specification, NUREG 1433.BFN-UNITS 1, 2,&3 Revision 0 0 t-JUSTIFICATION FOR CHANGES CTS 3.6.B/4.6.B -COOLANT CHEMISTRY'ELOCATED SPECIFICATIONS Rl The chemistry limits are provided to prevent long term component degradation and provide long term maintenance of acceptable structural conditions of the system.The associated surveillances are not required to ensure immediate operability of the reactor coolant system.Therefore, the requirements specified in current Specification 3.6.B/4.6.B did not satisfy the NRC Final Policy Statement technical specification screening criteria as documented in the Application of Selection Criteria to the Browns Ferry Unit 2 Technical Specifications and have been relocated to plant documents controlled in accordance with lOCFR50.59. Revision 0 PAGE JUSTIFICATION FOR CHANGES CTS 3.6.G/4.6.G -STRUCTURAL INTEGRITY RELOCATED SPECIFICATIONS Rl The structural integrity inspections are provided to prevent long term component degradation and provide long term maintenance of acceptable structural conditions of the system.The associated inspections are not required to ensure immediate operability'f the system.Therefore, the requirements specified in current Specification 3.6.G/4.6.G did not satisfy the NRC Final Policy Statement technical specification screening criteria as documented in the Application of Selection Criteria to the BFN Unit 2 Technical Specifications and have been relocated to plant documents controlled in accordance with lOCFR50.59. BFN-UNITS I, 2, 8L 3 Revision 0 peCiE~OF~'

JUSTIFICATION FOR CHANGES CTS 3.6.H/4.6.H -SNUBBERS TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LAl Snubber inspection requirements are part of the BFN Inservice Inspection (ISI)Program and are being relocated to the ISI program documents. Requirements for the ISI Program are specified in 10 CFR 50.55a to be performed in accordance with ASIDE Section XI.NRC regulations contain the necessary programmatic requirements for ISI without repeating them in the proposed BFN ISTS.Changes to the ISI Program are controlled in accordance with 10 CFR 50.59.With the removal of operability requirements from the Technical Specifications, snubber operability requirements will be determined in accordance with Technical Specification system operability requirements. BFN-UNITS 1, 2, 5 3 Revision 0 Section 3.4, Reactor Coolant System (RCS)Bases The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content of the proposed Browns Ferry Unit 2 Technical Specification Section 3.4, consistent with the BWR Standard Technical Specification, NUREG 1433.The revised Bases are as shown in the proposed Browns Ferry Unit 2 Technical Specification Bases.BFN-UNITS 1, 2, Ea 3 PAgp evi sQF~

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP C 1 4 0 LI TIHC COHDITIOHS FOR OPERATIOH SpE'Cgi QLQo.,)NOV 22l988 SURVEILLAHCE REQUIREMEHTS 3.5 4,5 0 G Applies o the operational status of the core and containmen cooling systems.Appli to the surv llancc reqair cats of the rc and containm t cooling s terna when the corre pending limi ng condi-tion for o cration is i effect.To assare the OP ILITY of the core and conta ament cooling systems under all c nditions for which thf.s cooling c pability is an essential respons to plant abnormalities To verify the PERABILITT of the core and contai cnt cooling systems under al conditions for which this cool capability is an essential response to plant abnormalities. L,CD 1.The CSS shall be OPERABLE: 3iS.l (1)PRIOR TO SI'JLBTUP from a COLD COHDITIOK> or (2)when there is irradiate fuel in the vessel and when the reactor vessel prcssure is greater than atmospheric prcssure, except as specified in Specification 3.5.h.2 Ll SR ax.l"i a-rage CC~'4+SR3eS I J C IlI Simulated Automatic Actuation test~i~u~c Once/I8~, Oyera&ag SR3~6b.Pump OPERA-BILITY Per Specifi-cation 1.0.MM Co tor Per Sycc+i-Op rate atioh, l.~Val e OPE ILI 1.Core Spray System Testing.5g P,S,],6 d.System flow Once/W rate: Each loop shall dclivcr at least 6250 gpm against a system head corres-ponding to a BFH Unit 1 3.5/4.5-1~AMENDMENT NO.15 9 FA+', sF t5 I S fcc,'0'c~k'on 3.S.l AUG 02 t989 5'4 B.S.i.4 105 psi differential prcssure bctveen the~reactor vessel and the primary containment. 2~+Cog)a)Aires csc e OPERAB rs gL/Sc>'n/Node 3 in Iahr If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days rov/t5 a ac ve components in the other CSS loop and thc RHR stem c hal V ve er ccif cati n MM S'g P f l.2..Once/Verify that each valve (manual, povcr-operated, or automatic) in the infection flovpath that is not locked, scaled, or other-vise secured in position, is in its correc g7 positione 3.If Specification 3.5.A.l r Specification 3.5.A.2 c ot RCIToN bc met, the reactor shal be 84 H'laced in the COLD SHUTDOWH COHDITIOH vithin ho s.S6 ca 4.Shen the reactor vesse pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERABLE pump and associated diesel generator shall bc OPERABLE, except vith the reactor vessel head remove s ccific as spccificd in 3.5.A.1.2~Eo Idigio s+gl~e r ited Sc'c X~l'*echo~ ger Agre'>>SPH tsTS S.t.l Except that an automatic valve capable of automatic return to its ECCS position vhen an ECCS signal is prcscnt may be in a position for another mode of operation. BFH Unit 1 SCe 34SHCfca4o~ P,~fjgggS BvH isis y,s.~3'/4,5-2 PAGE~OF~hMENDMENTNO. 16 9

The RHRS shall be OPERABLE fP.(1)PRIOR TO STARTUP from a COLD CONDITION; or 2)when there is irradiated fuel in the reactor vessel and when the reactor vessel prcssure is greater than atmospheric, except as specified i'Specifications 3.5.B.2 ou h 3.5.B I.a.SRZ.S.t S b.C~GRAS.I 4 d-L.l AI4l oi Simula ted Automa tie Actuation Test Pump OPERA-BILITY Motor Opera-ted'valve OP ERAB ILITY Pump Flow Rate O ce/,s 3 Per Specification 1.0.Per Specification 1.0.MM Once/9 mes4hc 5g35 i g fO Test Check Per Valve Specification l.O.MM c/e Verify that cc each valve (manual, power-s operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, ig i correct position.Zld~w'R 3.5.l.g/4oK SR3,.S,I.V g.Verify LPCI ce/subsystem cross-tie valve is closed gull power removed f rom valve operator.BFN Unit 1 Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5W Ex ept hat an a orna c val e apab of a to-mati retu to its ECC posit on w en an CCS s gnal is pr sent y b in a posit n f r an her ode of o e tio AMENOMENT gg, 2Pg PAGE ~k'cia'~Hen P,S;/AUG 02 tGGG 2~Vith thc reactor vcsscl pressur less 105 psig, the may be emoved from s rvicc (cept hat tvo RHR p ps-cont en cooling mode d asso atcd eat cx ers t r ia OPE)fo a pe iod not to exceed hour vhile b dra ed of s press n cham er qu ity vatcr fille vith primary coolaa quali y vater rovide that ring coold vn tvo oops th o e pump er lo or o loo vith tvo p ps, an associate diesel generatorsy ia e core ray stem are OPE s RS.XI Each LPCI pump shall deliver 9000 gpm against, an indicated system pressure of 125, psig.Tvo LPCI pumps in thc same loop shall deliver 12000 gpm agaiast an indicated system pressure of 250 psig.2.An air test oa the dryvcl aad torus headers and nozzles shall be conducted once/5 years.A vater test may bc performed on thc torus hcadcr in lieu of the air test.Se'e 3usRFicogon P<QQQC'5 gi BFN~5'4'ig 3.If e RHR um (LPCI mode)is inoperable, the reactor~may remain in operation for a period not to exceed 7 days AcT>NJ H r cma RHR pumps (LPCI mode)and both access paths of the RHRS LPCI made)and the CS PERABLE.5 CC Su5AQ<a~~At Ch~g&~BFu tSTS Z,S,l Ll 4.If 2 RHR um s (LPCI mode)become inoperablc, the reactor shall bc placed in the CO SHUTDOWH COHDITIOH vi thin hours.3'c.Qei Jn Noh'ia l2hrs Rnol BFH Unit 1 3.5/4.5-5-a.-L~ne~g I NENDMENTNO. 16 9 0 Qpccl gicahon 3,5;)8.ACT loQ5 B 4H.If Specifications 3.5.B.1 thro h are t an e reac or shall be placed in the COLD SH WN CONDITION within hours.Pb LZ, When e pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. inMQs 3"~5~q Sec'&Cicahora~M~S 4w 8p J4 l5TS 7>>$.%9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.10.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling re not required 10.No additional surveillance required.BPN Unit 1 When there xs irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capabi.lity is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5>>7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.NEHOMENT NO.204 PAG~

1989 12.If one RHR pump or associated. heat exchanger located the unit cross collILectio the ad)a t unit is i operable fo any reason (i eluding val e inoperability, pip break, etc , the reactor may emain in op ation for a eriod not exceed 30 day provided remaining RHR pum and associ ed diesel generato are OPE 13.If RHR cro s-connection lov or heat remov capability lost, the unit may remain in ope tion for a period t to exceed 10 days unless su capability is btoredo 12.Ho additional surveillanc required.13.Ho additional surveillance re uired.SC B.S..14.1 reci culati n pump di charge valves shall be P PRIOR TO S (or close if permi ted el ether in the speci cations)14.All recirculation pump discharge valves shall be tested for OPERABILI. during any period of gg p g (,S COLD SHUTDOWH COHDITIOK exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.BFK Unit 1 3.5/4.5-8 AMENDMBlr 80.y 69 PG 7 oF Is SF'c;'on 3.5.I 5 c r~s+,'<'mt'on Q~C4+eS@r l~TS.St.cHon 3.3.g.(St.4 Sus+4 ta fjon 5a~c's foe BFn)lSTs 3.V.7 4.9.4.4.(Cont'd)Ce d.'4-kV a own board voltages shall be recorded once every 12 hours.The loss of voltage and degraded voltage relays vhich start thc diesel gcncratora from the 4-kV shutdown boards shall be calibrated annually for trip and reset and the measarcmcnts logged.These relays shall be calibrated as specified in Table 4.9.k.4.c. 5.Logic Systems a.Coamon accident signal logic system is OPZRhBLE.>NSti Acegy~P g Qe BC'N)Sy>Z,~,l 5.480-V RMV Boards ID and 1E SR 3,5.l.l?LCt~i%ao Once c 0$e automat c transfer feature for 480-V RNOV boards ID and 321 shall be fanctionally tested to verify auto-transfer capability. b, 480-V load ahedd ,logic system is OPERJUKE.6.There shall be a minimum of 35,280 gallons of diesel fuel in each of the 7-day diesel-gcncrator fuel tank assemblies. 5~<~+if':nk'en fv Chr~X+A)f5'Q j f g BPS Unit 1 3 9/4.9-7 NENOMEHT Ng.y 8 g PAGE oF I~ $Pcc,i4'.5'.1 NOV 0 4 199t 3.9.h.d.The 480-V shutdown boards lh and 1B are energized. See Yustihcation far Changes 5r at=~1sT',g,7 e.The units 1 and 2 diesel auxiliary boards are ized f.Loss of voltage and degraded voltage relays OPERhBLE on 4-kV shutdown boards h, B, C, and D.g.Shutdown buses 1 and 2 energized. ~Ce Su544icaHon Par f/~&so~isis z.z.Li 5CC ScaS+kcce'gase Qg Q/IongeS'~ BFN isaac p,g,~h.Th-react r mo or-op rated valve RMO b ards&1E are ener ized, th m tor-ge era r ()ets I, 1D>and lEh se ice 4.The three 250-V unit batteries, the four shutdown board batteries, a battery charger for each batte an assoc ated battery board are OPERhBLE.See g~gqg;cab'on f r Qaga W BPH 1575 3.f.g a~d'tt.7 4.Undervoltage Relays a.(Deleted)h.Once every 18 conchs, the conditions under which the loss of voltage and degraded voltage relays are required shall be simulated with an undervoltage on each shutdown board to demonstrate that the associated diesel generator wi'll start.PAGE BFH Unit 1 3.9/4.9-6 AMEHOMEHT tbtO I 8 6 NOV 18 1888~I 12.When one 480-V ahutdovn board is found to be ZEOPERhBLE, the reactor vill be placed in HOT STANDBY COHDITIOS vithin 12 hours and COLD Q93TDO OIITI01 vithin 24 hours.13.If e 480-V ard ag se is ZEO REk R 0 05 tinue for a eriod t o exce sev days proteid the eaa 480-V ard sets and ir so cia lo 0 4.Zi aag 480-RMV aS sets ecoa Z50 rea tor shall pla in the COLD S CO TIO vithin 24 urs.5.If the reqaireaents for operating, in the coaditioas specified by 3.9.B.1 through 3.9.B.14 cannot be aet, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWNS COHDITIOK vithin 24 hours.See X~5+pjc'Phon 5c Chants fee l5 TS B.f 7 See 345 hPicggon for (4~gy t))-"H)S)S Sac@an).E BFN Unit 1 3.9/4.9-14 PAGE Cn pP l5 AMENDMENT NO.y5 8 SPCC 4'eA 3,5, FEB 0 7 I99I 3.5.D 0 s 4~5.D 1.The equipment area co er asso ated vith each pump the equipment area'coo r associated vith each t of core spray pumps and C or B and D)mu be OPERABLE at all mes en the pump or p s se ed by that speci c coole is considered t be OPE LE.l.Each e ipme t area coole~is opera ed i con)unction vith the quip nt served y that pa ticu r cooler;erefore, e e uipment a a cooler are ested at the same freq ency s the pump vhich th se e.2.When an equ ment area cooler, is not PERABLE, the pump(s)se ed by that c er must be co idered inop ble for Tec ical Specific on purposes.c.co 1.3.s.(Wr'f('CR4l ~7 (ll]P(ep5eck~ok 4~$R3,5,(,g The HPCI sy em s all be enever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWH COHDITIOH or as specified in Specification 3.5.E.2.OPERABILITY shall be deter-" mined vithin 12 hours after reactor steam pressur (oA reaches 150 psig from a COLD COHDITIO r a ernat ve y RI TO ST P usi an 1 ry ste su ly.M3 S.(<r a.copse(hloR br sR3~(,'l c~HPCI Subsystem testing.shall be.performed as follovs:+fu or Simulated Once/18 Automatic months Actuation Test 3 Pump OPERA-BILITY Per Specification 1.0.Moto Oper-Per ted alve Spe fica$io 0 RAB ITY.0.sR 3,g.l,g r~(Oow SR s.s,(.q Flov Rate at Once/4-a monQm ra or e PI3 oe tng pr s re$2o&lolo(s,'g lg BFH Unit 1/t(Q 3.5/4.5-13 AMENDMEHT NO.I S 0 Spec'4'<<a< 3',s, (FEB 0 7 1991 Coolant In e 2~~Bog//<~re/Qfioa p LPCI an OPERABLE.S are VCce P~till~~If the HPCI system is inoperable, the react r may remain in operation a d not to exceed M ays, prov ed t e R Llo K/4S+o'p SR 3.5, i.w sR z.s.>.8 Flov Rate at Once/18 psig months Verify that each valve (manual, pover-operated, or automatic) in the injection flow-path that is not locked, sealed, or othervise secured in Once R3 position is in its correc osition.W The HPCI pump shall deliver at least$000 gpm during each flow rate test.ld~s 3~If Specifications 3.5.E.l~~or 3.5.E.2 are not met, t e reactor vessel pressure~ac s shall be reduced to 150'"'sig or less vithin R4 hours.3&*E cept that an aut matic va ve c able aut mati retu to s ECC posit on wh an ECCS ignal is pre ent may b in a ositio for anothe mode o operation. F.'eacto Co e splat o Coo in F.Reactor Core Iso at o Cooli~~~BFN Unit 1 1.The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall e~X~A;Rc ho<4c'~~5.5/4.5-14 O'C 80%lSYS SaSo3 1.RCIC Subsystem testing shall be performed as follows: a.Simulated Auto-Once/18 matic Actuation months Test NIENOMEMt Hp.X 80 pAQF/~QF~~~

SgeCig'ro,>on g, g.NY 1 9 l994 Qqhcqb's)iQ Six valves of the Automatic Depressurization System shall be OPERABLE: (1)PRIOR TO STARTUP from a COLD CONDITION, or, L ISb (2)whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 485 ps g, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.PAfaScd 4~4 Sg g,S,ll f he relief gal s is ove d 4.6..2)9$'.Durin ea cretin c he following tests shall be performed on the ADS: L(Ac~i Of SR P,g J lg a.A simulated automatic actuation test shall cut~l8~55.be performed PRIOR TO STARTUP te each outa 2~P'GTlodb E WTlog 8+8 3~R CTiDA5 G.With one of the above required ADS valves inoperable, provided the HPCI system the core spray system and the LPCI system re OPERABLE, res ore the inoperable ADS valve to OPERABLE statue within 14 days or be in at least a HOT SHUTDOWN CONDITION within the next 12 hours and reduce reactor steam dome ressure to~495 psig within hours.IS@sc With two or more of the above required ADS valves inoperable, be in at least a HOT SRJTDOWN CONDITION within i2 hours and reduce react steam dome pres'sure to g+95 psig within hours.ISo 3'4, L2, al recpaka ed-.frig s+c(Sg 3,5/3 frofasect AcTloH p Rss~'rM Ps@on L)(LCo 3.s,gg QopB 2 cubi whig 7gry Old DE 3 QPifhirt (3/r PlODE 9 luis n 3'7/lfS LI L I's (W A~<nlrb)BFN Unit 1 3.5/4.5-16 NENOMENT NL 2P g PAGE~3 OF~5 7 S eCigca.'on R,g 0KC O V 894.C 6~C 2.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours I for the air sampling system..!!The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.See Y~sSka Hen P, Cg~~W BPt4 ls75 g.q,g 3.If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.4.e.D 1.When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours.The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. 1.Approximately one-hal of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.All 13 valves will have been checked or replaced upon the completion of every second cycle.+~>4"<wagon 4c CQ~4 SPA)sTs g~,~Pfop6c'd~a4~S.S.i.U LR 2.In accordance with Specification 1.0.MN ach elie va ve shall be manually opened unt ermocoup es and a usti monit s do stre of t e val e ind cate earn i ow fro the ve.BFN Unit 1 3.6/4.e-lo AMENPlHEHT NL 2 Z3~3 PAGE 1 QF I5 Whenever the core spray s stems, LP I, HPCI, or C ired to e OPERAB , th discharg pipi from the pum discharg of the e systems to t e last block v lvc shall be f led.Rig~SC S.S.<,1 e following surveillance requirements shall be adhered to assure that the discharge piping of the core spray systems, LPCI~HPCI, d RCI are filled: Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 8 psig 48 psig S~c~~f'cabin f c Chavez)a f4'~~)STS 3,g.y The suctio of the IC an pumps shall e aligned to the condensate storage tank, and the pre ure supp ession chambe head tank shall no lly be aligned to erve the discharge piping of th RHR and S pumps.Th condensate head t may be used to serve th RHR an CS disc ge piping the P hea tank i unavailable The pressur indicators the discharg of the RHR CS pumps sha indicate not less than liste below.1.ve mont h he RHRS LPCI and Conta ament Spray)and core spray system, the dischar pipin of these s stem Pll v ed om e h'gh oin wa fl d rm ed 2.F ow ng any per od where the LP I or ore s ray s t hav not een r quire to e OP LE, he di charg pi ng of t ino rable syst sha 1 be v ted f m the high in prior o the eturn of the system o service.3.Whenever the HPCI or RCI system is lined up to take suction from the condensate storage tank, the dischar e pipin of thc HFCI an CIC bc v n p t f t s t ob rve on a monthly basis.4.en th RHRS and th CSS are r uired o be OPERAB , the pr sure x dica ors whx h mon or th disc arge li es shall be mon tore daily d the pressure recor ed.PAGE~OF L+BFN Unit 1 3'/4.5-17 AIENOlHENT No.2 05 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP 5 ce Sic 4io~'3~~4 0 0 COO G S S S p)NOV 22 1988 LIMITIHG COHDITIOHS FOR OPERATION SURVEILLAHCE REQUIREMENTS 3.5 CO MSHXILS CO COO G 4.5'O CO~SS~~S COO G cab cab t A lies to the operational st us of the core and coat ament cooling systems.Applies to the s eillance requirements of th core and containment cooling ystems vhea the corresponding lim ting condi-tion for operatioa is effect.0 ect ve 0 ect vc To assure the OP ILITY of thc core and conta cnt cooling systems under all co itions for vhich this cooling cap ility is an essential rcsponsc to lant abnormalities. To ve fy the OPERABILITY of the core an containment cooling systems er all conditions for which this ooling capability is an essential zesponse to plant abnormalities. ca o Sec cto t ahe 3.5.l (1)4ppl;((4'.Ig <>>CSS shal1 be OPERABLE: PRIOR TO STARTUP from a COLD CONDITION~ or rhea there is irradiated fuel in the vessel and when thc reactor vessel pressure is greater than atmospheric pressure, except as specified in Specification 3.5.A.2.sR 3.s.l.g a.P,l gQ tle4,*sea.s,l.t act Ao lated Automatic Actuation test g~v~enc cc]e Qpomtekag p"t PRIES I f P3 Pump Opera-Pcr Specifi-bility catioa 1.O.MM c.otor O Op ated pg Valv OPERABILITY Per pec fi-n~en l.h MN 1.Core Spray System Testing.san.s.l.(, d.System flov rate: Each loop shall deliver at least 6250 gpm against a system head corres-ponding to a Once/M~P'~/qz.4~s BFH Unit 2 3.5/4.5-1 AMENDMEHTNO. 15 5

ON g gC,,'fir.qC(o~ ~<I~'S Cont'd)g.s.t.6 105 psi differential pressure bctveen the reactor vessel and the primary containment. P~e.Ch ck Valve Per Speci cati a.o.m 20 PA'ld 9 If one CSS loop is inoperable,.the reactor may remain in operation for a period not to exceed 7 days rovidi all active components in the other CSS loop and thc PCI mode and the diesel generators are OPERABLE.Sc w Ho~IR Qrg Once/each valve (manual, pover-operated, or As automatic) in the~ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc p7 position.3~Ad riOat~+8 4, If Specification 3.5.h.l or Specification 3.5.h.2 cannot be met, the reactor shal e placed in the COLD SHUTDOWH COHDITIOH vithin hours.LQ.When thc reactor vesse pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERJQKZ pump anil associate4 diesel generator shall be OPERABLE, ezcept vith the reactor vessel head remo c ln 3.5.A.or PRIOR TO STARTUP as specified in 3.5.h.l.OA6 addit nal s~eQ lance is r ire 74 s4iflc44s~~Col4<<wg+J g4~a~~(srs~.s.~ccpt that automatic v ve capable f autom tic re rn to its CS posi ion vhen ECCS sign is presen may be in a position or another mode of operation. BFH Unit 2+~~~>4 t,'cA(d~4yi Qd~*-gS/Is~~3.5/4.5-2 PAGE~QF~5 AMENDMENT NO.16 9 oo xng nme t ent)PRIOR TO STARTUP from a COLD CONDITION; or 2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.B.2, through 3.5.B.7.The RHRS shall be OPERABLE 0.Simulated Automa tic Actuation Test l.a.ce>g Opee~Ig Oyer c.otor pere-er ted valve Specxf xcatxon OPERABIL Sg S.5,).C d.Pump Flow Once/&Rate 6 e.Testa e Check A~Valve Per Specification 1.0.MN S~~<l g b.Pump OPERA-er BILITY'pecification 1.0.MM 5R 35I.'L f Once/04m+h Verify that each valve (manual, po~er-operated, or J/cg automatic) in the injection flow-path that is not locked, sealed, or other~isa secured in pcsi-tion, i in its P'7 correc positron.81 Unit 2 3.5/4.5-4 sR 3.</.2.So@.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. g.Verify LPCI subsystem cross-tie valve is closed nruj power removed from valve operator.nce/AMENPMgfT gg.2 2 P Except t>at an automat c valve capabl of auto-mati return to'ECC position en an CCS signa is present may ve in a position for another mode of operation.

S ci fic~A~w 3.5.I AUG 02 5gg 3 4~nt ainment 2~With the rea tor vessel pressure le than 105 psig, the RHRS'be removed from serv ce (except that tvo RHR pump-containment cooling mode associated heat exch ers must remain OPE LE)for a per d not to ceed 24 hours hile b ng drained of ppression ch er quality vater and fill d vith primary cool t quality vater provi ed that dur cooldown o loops wi one pump per oop or one oop vith tvo pumps, an associated diesel generators, in th core s ra s stem a LZ 0 0~)sR 3.S.ach LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Two LPCI pumps in thc same loop shall deliver 12000 gpm against an indicated system pressure of 250 psig.2.An air test on the dryvell and torus headers and nozzle shall be conducted once/5 years.h vater test may be performed on the torus header in lieu of the air test.c See: a'aA4<J>o<g~~imam r.C.2 3~PcTI0 hl A p,cgw H Lf If one RHR ump (LP mode)is inoperable, the reactor may remain in operation for a period not to cxcecd 7 days rov e e rema n ng pumps (LPCI mode)and both access paths of the LPCI mode an S an t c cse generators remain OPERABLZ Pl~S,ea-s$:AcJ o-4'-Cw~(~~Sir BFN la~<.S.l e l~4.If 2 umps (LPCI mode)become inoperable, the reactor shall be placed in thc COLD SHUTDOWN CONDITION vi thin hours.3b$2.plo6 3 is./2 4rS ance BFN Unit 2 3 5/4.5-5~(.-E~OF lP AMENDMENTNO. l6 9

m Pi 4 tainmen t inmen t 8.Acvlo~l2 6w nce Sce~~s]A'zi4~ 4r C4o~igg ger Bf'N<S~)Rr.~esse pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.0.If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required 10.No additional survei'lance required.Unit 2 When there is irradiated fuel in the reactor and the reactor is'ot in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capabili.ty is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.J AMENDMENT HD.2 2 3 5'c i'gic<4io AUG 02 198Sent va S 12.If three RHR pumps or associated heat exchangers located on the unit cross-connection in the ad)acent uni s are operable for any r son (ncluding valve inopc bility, pi e break, etc.), thc r ctor may emain in operation for a eriod not to exceed 30 day provided the remaining RHR pump and associated diesel generator are OPERABLE.3.If RHR cross connection flov or heat removal pability is lost, the unit may re ain in operation for a period not o exceed 10 days unless such c ability is restored 4.11 rec rcu at on ump d charge vcs sh ll bc ERABLE IOR TO STAR (or cl scd if permit d clscv re in these specifi tions).13.No additional sur illance required.gg 3.5.(,5 14.411 recirculation pump discharge valves shall bc tested for OPERABILITY uring any period of sq q~~~COLD SHUTDOMH CONDITION exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.Unit 2 3.5/4.5-8 AMENDMENT NO;1 6 9 WGE TZOH A>S ec',f<e 4'.S.NOV 0~1991 HTS 4 d.The 480-V shutdown boards 2A and 2B arc energized. e.The units 1 and 2 diesel auxiliary boards are energized. See~~xlP+lw cog r g-gem ic.rs 3.8.7 f.Loss of voltage and-degraded voltage relays OPERABLE on 4-kV shutdown boards A, B, C, and D.g.Shutdown buses 1 and 2 energized. The 480 reactor motor-opera d valve (OV)boar s 2D&2E e encr ized wi motor-gen ator g)s s 2DH, 2D , 2EH, d 2EA service.see~~sligicPaw 4l c4~Jag~PpN Ic7s 2.3.8.J S e~us4Ki~k.o. 4r n.-g~4<NFL Isis 3.8.J The three 250-V unit batteries, the four shutdown board batteries, a battery charger for each battery and ssociatc atte board arc OPERABLE.4.Undcrvoltage Relays a.(Deleted)b.Once every 18 months, thc conditions under which the loss of voltage and degraded voltage relays arc required shall be simulated with an undervoltagc on each shutdown board to demonstrate that the associated diesel generator will start.3.9/4.9-6 AMENDMgfTN0,$9 g 4.9.h.4.(Cont'd)QQ, 3~gle ksc41>~phag~Qr LJ~f5+/5CC4 l~~3 8'I cc 4~L4i4<ce4 4t C~~QI-gP'tJ J S T'5 3.8.7 c.The loss of voltage and degraded voltage relays which start the diesel generators from the 4-kV shutdova boards shall be calibrated annually for trip and reset and the meaaurcmenta logged.These relays shall be calibrated aa specified.h.4.c.4-kV shutdown board voltages shall be recorded once every.12 hours.5 Logic Systems a.Comaon accident signal logic system ia OPERABLE.5.480-V RNOV Boards 2D and 2E I 5~is.~P e aut c transfer feature for 480-V RMDV boards 2D and 2E shall bc functionally teated to verify auto-transfer capability. b.480-V load shedding, logic system ia OPERAS.6.There shall be a miniaaun of 35,280 gallons of diesel fuel in each of the 7-day diesel-generator fuel tank assemblies. See r~k l;o<~.Ch.Zc~~gin irrs R,Z.3 BES Unit 2 3.9/4.9-7 NEHDMEHT NO.1.9 I.)s L.Whea onc 480-V shutdovn board is found to be IHOPERABLE, thc reactor vill be placed in the HOT STAIBY COHDITIOH vithin 12 hours and COLD SHOTDOWH COHDITIOH vithin 24 hours.13.If onc 48 V RNOV board mg set is I PERABLE, REACTOR POWER 0 ERATIOH may cont e for a period not to cccd seven daysg pr ided the rema 0-V RMOV board sets and their associa ed loads remain OHGKBLE~I 4.If auy tvo 4 0-V RMOV board mg s s become IHOPE , the rea or shall placed in c COLD S WH CO ITIOH vithin 24 hours.15.If thc requirements for operating in the conditions specified by 3.9.B 1 through 3.9.B.14 cannot be mct, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWH COHDITIOH vithin 24 hours.+C ZML4i C+g~~~h~g 4av QpW I S7 5 5~~Qy~p g BFH Unit 2 3'/4.9-14 AMENDMENT NO.Zg 4 pAGE 0 OF f5

Cl~: livia(g t~Iitl I~~~~~II~~~~~~~I~~~~II~~~~~~~~t I~I~I~~~~II~~~~~~~~~~~~~I~I I I~~~I~~~~~I~III~~~~~\~I t~~II~~~~~~'~~~~~II~~~~~~~I~I~~~~~'I~~~~~~~~~rr'I'MrttE tr 4 I t O'3 I 8 0 4+0!I%~ax r m i I~I II t II~~~r I~~~~~~~~~~~I~~~~I~~~~~~I~~~I I'h~~I~~~~~~~~~~I~~I I~I~'~'~I~~I~.~'~~~~~II~~~>~r ClS' 5'iVi'cc4io~ 3.R/FEB 0 7 t9gt ct on Io c (45 P sip SR S~Flov Rate at Once/18 448-psig months The HPCI pump shall deliver at least 5000 gpm during each flov rate test.SR f~E.C.I.2 Verify that Once/each valve (manual, pover-operated, or automatic) in the injection flov-path that is not locked, sealed, or othervise secured in positio , is in its correc position.Pro Ahviog D 2~If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed I days, rovided the RHRS(LPCI and RCICS are~OP LE vtr nc pcyio~As~3~p,n iota g4H If Specifications 3.5.E.l or 3.5.E.2 arc not mct, t e~o PE reactor vessel pressure shall be reduced to 150 psig or less vithin 3'ours.*ept that au automatic va e capable o autom tic retu to its EC posi on vhen a ECCS signa is present y be in a position or another mode of operation. a o o F.a t C 0 oo BFN Unit 2 The RCICS shall be OPERABLE vhenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 3.5/4.5-14 1.RCIC Subsystem testing shall bc performed as follovs: a.Simulated Auto-Once/18 matic Actuation months Test AMENDMEHT NO.1 9 0 Sec Z~>$;~iicqfjo~ 4~C~ra egal]g7g 5 5'3

FEB 0 7 199t 1.Six valves of the Automatic p,g Depressurisation System shall be OPERABLE: (1)PRIOR TO STARTUP from a COLD COHDITIOH, or, I p)~Q~(2)vhenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than si except in the COLD SHUT-DOWH COHDITIOH or as specified in 3.5.G.2 and 3.5.G.3 belov.(Vg~+(5~rg.Pro~Ny W s~3.s.I.LQ rve~cc o the rel val&s is vercd i 4.6.D.1.Duri each operat c c 0 ov ng tests shall be performed on the ADS 4'R a.A simulated automatic 3 5.(lO'ctuation test shall be performed PRIOR TO STARTUP er eq efuel c 2~P,neo&jEcTlo 4 e~H Pith one of thc above required ADS valves inoperablc, provided the HPCI s stem, e core spray system and the LPCI stem re OPERABLE, restore the inoperablc ADS valve to OPERABLE status vithin 14 days or bc in M least a HOT SHUTDOWH CO5QITIOH vithin thc next lg'nours and reduce reactor team dome pressure to sig vithin hours.ISO e ances P~~~sf 3.s;],g 9 o(-+Ace(ou r=3~Pcyso nE&Vith tvo or more of the above required ADS valves, inoperable, be in at least a HOT SHUTDOWN COHDITIOH vithin 12 hours and reduce reactor team dome pressure to sig vithin hours.58 b L.5 BFH Unit 2 R4'g~~M Ackiy-g.I (C Z u 3.~)Robe 2-1,~'7 ggg vf<I34 s A J7 I rs~g.lZ 3.5/4.5-16 A DEC 0 7 l994 G COND 0 S OR 0.6.C Coo 2~CC I I I4 I age Anytime irradiated fuel fs in the reactor vessel and reactor coolant temperature fs above 212 F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, thc reactor may remain in operation durfng the succeeding 24 hours for the sump system or 72 hours for the afr sampling system.4.6.C Co a a 2.With the afr sampling system inoperable, grab samples shall be obtafned and analyzed at, 1cast once every 24 hours.S~<+'S Ja 4 C J.;O~CO/'M gag fo~SPV'i<7-S~.q.g The afr sampling system may be removed from service for a period of 4 hours for calibration, function testing j and maintenance vithout providing a temporary monitor.3~If the condition in l.or 2 above cannot be met, an orderly shutdovn shall be initiated and the reactor shall be placed fn the COLD SHUTDOWNS COHDITIOS vithfn 24 hours.4s6 D BPÃUnit 2 20 Sec~,t;g;<;P,C<~W QFIIJ IS~P.Q.3 I R'I.S.I.L I OJo4.~Sa 3.s.l.II c8 3.6/4.6<<10 When more than onc relief valve is knovn to bc failed, an orderly shutdovn shall be inftiated and thc reactor 4eprcssurfzed to less than 105 psig vithin 24 hours.The relief valves are not required to be OPERABLZ in the COLD SHUTDOWN Approximately one-half o f all relief valves shall be bench-checked or replaced vith a bench-checked valve each operating cycle.All 13 valves vill have been checked or replaced upon the completion of every second, cycle.In accordance vith Speci cation 1.0.rclicf valve shall be manually o ened unt ermo coup es ac tic monitors do earn of valv indicat steam is f loving from thc valve.AMENbMENT NO.2 29 P GE~

5 ci4ica4iom 3.5.(AU 0 1989 3.5.4~~~s&e Whenever the core spray systems, LPCI, HPCI, or RCIC required to be OPERABLE, the disc pipi from the pump disch ge of th e systems to the las block lve shall be filled.~a S g 3.C'.l.1 The folloving surveillance requirements shall bc adhered to assure that the discharge piping of the core spr systems, LPCI, HPCI, d RC C are filled: g.A 7 The sucti of the IC an pumps shal be aligned to the condensate s orage tank, and thc pressure ppression chamber head tank shall ormally be aligned to serve the dis arge piping of th RHR and CS pum s.The con sate head t may be used to s e the RHR and S discharge pipin if thc PSC head tank is unav lable.The pr sure indicato on the discha e of the RHR and CS umps shall in cate not less th listed belov.Pl-75-20 48 psig Pl-75-48 8 psig P1<<74-51 48 psig Pl-74-65 48 psig""<~~~<bio gr Z4w-~$04 gag IgyS 1.Every month t e RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems sha 1 e volte rom t hig po~t and vat flov detc ined 443 2.Fo ov ng any period vhere t LPCI or core spray systems hav not been r uired to be OPE LE, the di harge, iping of th inoperable stem hall be vent from the igh point prior to the return of the system to service.LA9'.Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage tank, thc discharge piping of the HPCI and RC s c ve e rom the gh poin f thc s stem had vater flov obse d on a monthly basis.4.When the RHRS and the CSS are r uired to bc PERAB the pre ure indicat vhic monit thc dischar lin shall be monitored daily and thc prcssure recorded.OBFH Unit 2 3.5/4.5-17 AMENDMENT tltO.I 6 9 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 8 CO COO~z.i MQV 22 1988 4.5 MIXER 00 G kppli.es the opcrationa status of e core and containmcnt cooling systems.lies to the surve lance rcq rements of thc co e and conta ent cooling sys when thc cor cspondipg limiti condi-tion for peration is in e feet.To assure the 0 LITT of the core and containm t cooling systems under all cond iona for which this cooling capab lity is an essential response to ant abnormalities. To veri the 0 ILITY of thc core and conta t cooling systems under all onditions for which this cooling pability is essential response to plant abnormalities. t aco 3 5'-((1)PRIOR TO STARTUP from a COLD COHDITIOK, or Ay l'cab4)(2)when there is irradiate fuel in the vessel and when the reactor vessel pressure is greater than atmospheric pressure, except aa specified in Specification 3.5.k.2.1.The CSS shall be OPKHBLE: SR s.s.).g Pn~sed]u.>+see.s..ao b.Rg sg p.g,>,L c+Simulated Automatic Actuation teat Pump OPBM-BILITX 9 Once/Q~Q&44LC441g 4p6kc Pcr Specifi-cation 1.0.c.~otor 0 eratcd V ve ILIA r Spe ifi-ca on 1 O.MM l.Core Spray System Testing.sZZc.l.(d.System flow rate: Each loop shall deliver at least 6250 gpm against a system head corres-ponding to a Once/&8Z m~?$BFH Unit 3 3.5/4.5-1 PAGE NENDMENTNO. t>O Ci +F1 f 105 psi'diffcreatial 3 5 J~(prcssure betvcea the reactor vehsel and the primary containment. 6c in hloge3 ZH I&bras 3.If Specification 3.5 h.l or Specification 3.5.k.cannot be met, the reactor shal be placed in the COLD SHOTDOWE COHDITI05 vithin hours.3C, L2, PC4'o nJ 8+H 2.If one CSS loop is inoperable, the reactor may remain in operation for a period aot to exceed 7 da rov ng 1 active components in iqC+ie~H the other CSS loop and th RHR stem CI mod e dicsc generator are OPERABLE.e.stable Per Che Valve Qpecifi tioa I.O.NN 5 3.5-.>Verify that each valve (manual y povcr-operatcd, or Bs automatic) in the inJection flovpath that is not locked, sealed, or othcr-vise secured in position, is in its oottsotg-Qyl position.additi~sotssiLLanos is ired.5cc~~gg;ecch'oit Qc change~r Se+ISV'S r,S, I A3 4.en the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop vith one OPERABLE pump and associated diesel generator shall bc OPERhBIZ, except vith the reactor vessel head remove as s ecificd in 3 h.r a specified ia 3.5.h.l Except that an automat c valve capable o autom ic r turn to its EC posit on vh aa ECCS signa is prcscn may bc in a position for another mode Scc gusttg~fio~+chcln)cf rr BAN ISTIC 3.5.2 BFH Unit 3 3,5/4.5-2 g,l Once/iB,s~aking Qcc1s-qua I~k Simulated Automatic Actuation Test RHRS shall be OPERABLE 8.The Q2.PRIOR TO STARTUP from a COLD CONDITION; or Applicab Iig (2)when there is irradiated fuel in the reactor vessel and when the reactor vessel pressure is greater than atmospheric, except as specified in Specifications 3.5.$.2 through 3.5.B 7.s83.5, l.0 Sg 3.S.I.Z oR PC~:~y~mme~~St~keff valve i~g~LPC(crust g,'t ts c4scd~C 3.S.l.Z Nsa Low pressure coolant injection (LPCI)may be considered OPERABLE during alignment and operation'or shutdown cooling with reactor steam dome pressure less than 105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. 3.5/4.5-4 BFN Unit 3&o b.c~e.Pump OPERA-BILI~er Specification 1.0.MM Motor Opera-Per ted valve Specification OPERABILITY 1.0.MM Pump Flow Once/W J Rate men ths-4ag Testable Per Check Specification Va,lve 1.MM Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion,'n its 7 correc position.Once/Verify LPCI Once/subsystem cross-tie valve is closed~power removed from valve operator.Except that an auto tic valve cap le of auto-ma ic retu to ts CS posi on en an ECCS igna is present may be in a position for another mode of operation. NENOMEMT HO.I 77 i 0 skci4 ca~n a.s.SURVZnumCE REqUIZZmm UB 02 2~3~+5'o 4 With the reactor csscl prc sure less t 105 psig, th RHRS may b removed f om service except tvo pumps-c tainmcnt ooling mode and a ociated h at exchanger must r OPEiULB for a pc od not to cxc d 24 hour vhilc being drained o supp cssion ber quali y va r and fi ed vith p ry coo t quali vatcr prov ded that d ing cooldovn o loops v th one pump pe loop or o loop vith pumps, associ tcd diesc generators, in e core spray system a e OPBRhBLE.1.If ne RHR um (I mode)is inoperable, the reactor may remain in operation for a period not to excccd 7 days prov e e rema pumps (LPCI mode)and bo access paths of the LPCI mode and the S and in a ors rema OPE~~~~R Each LPCI pump shall deliver 9000 gpm against an indicated system pressure of 125 psig.Two LPCI pumps in the same loop shall deliver 12000 gpm against an indicated system prcssure of 250 psig.2.hn air test on the drywell and tyne headers and nozzles shall bc conducted once/5 years.k vatcr test may be performed on the torus header in es See Zff5t)fi'Gabon &I Chases~c SP'iV (5 Tg g,g, g.q Sea$~,1'C'cJi gyral tsrS 3 f.I 4.If any'RHR um (LPCI mode ecome inoperable, the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.3C Bc,;r e&e3, in Izhrs BFH Unit 3 3.5/4.5-5 pAGE 5 oF~~hMENGMBlTNO. t a O Cl nt ent 8.Ron5 8+'If Specifications 3.5.B.l through 3.5.B.7 are not met,'a44keted d the reactor shall be placed in the COLD SHUTDOWN CONDITION within hours.3 R in~<3'~IPhrs A1 Sec JuSt 4>'capon Per Changes gr gP~ts rs s.s.~When e reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.The pumps'ssociated diesel generators must also be OPERABLE.Low pressure coolagt injection (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specif ication 1.0.MM.10.If the conditions of Specification 3.5.A.5 are met, LPCI and containment coolin are not requi d 10.No additional surveillance required.BFN Unit 3 When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 The B and D RHR p'umps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0;MM when the cross-connect capability is required.AMENDMENT NQ.g 77 pgsp~oF~t~ 8 AUG 02 t989 on 12.If one RHR pump or associated heat exchanger located on the unit cross-connection in unit 2 is i perable for any reason (eluding valve inopcrabilit pipe eak, etc.), the rc tor ma remain in operation for a eriod not to excee 30 days rovided thc remai RHR pump d associated diese generator a OPERABLE.3~14.I cross-co ction flov or heat emoval capabi ty is lost, thc un t may remain i peration for a pe od not to exce 10 days unless such capability is restored.h rec rculat n p ischarg valves shal b OPERAB PRIOR 0 ST (o closed f pe tted e cvhere in t se spe ficatio).13.Ho additional surveillance required.SR XS.14.All recirculation pump discharge valves shall be tested for OPERABIL during any period of gg 3~t~COLD SHUTDOWR COHDITIOH exceeding 48 hours, if OPERABILITY tests have not been performed during thc preceding 31 days.BFH Unit 3 3.5/4.5-8 AMENDMENT HO.I 4 0 PAG<~ S Stoic I g(ggt'EB i 4 1995 Sc<3'us,H f-:<chion Qe, ckclwy g I ST 5 Section 3'g (~ca>u~S'6;c tt'nnA, 0 ng c5 g~(et p~l STS Z87 4.9.A.4.(Cont'd)c.The loss of voltage and degraded voltage relays which start the diesel generators from the 4-kV shutdown boards shall be calibrated annually for trip and reset and thc measurements logged.These relays shall be calibrated as specified in able 4.9.A.4.c. d.4-kV shutdown board voltages shall bc recorded once every 12 hours.a.Accident signal logic system is OPERABLE.b.480-volt load s e logic system is OPERABLE.>ee 3<5(inca fjin far~go~'~(sTS S.g.(5.4-V V pard 3Rk2E a.Once sC r.s.i.u.c automatic transfer feature for 480-V RMOV boards 3D and 3E shall bc functionally tested to verify auto-transfer capability. 6.There shall bc a minimum of 35,280 gallons of diesel fuel in each of the 7-day diesel-gcncrator fuel tank assemblics. 5ea$wy;p;a.4on Qr~~Ri pFg BPS Qnit 3 3~9/4~9-7 AIItEHDMEÃf Rtt.I 8 9 PAGE 8'oF~~ 8 , NOV 0 4 1991 e.Loss of voltage and degraded voltage relays OPERhBLE on 4-kV shutdown boards 3Eh, 3EB, 3EC, and 3ED.f.Thc 480-V diesel auxiliary boards 3Eh and 3EB are encrgizcd. g.The 480-V reactor tor-opera d valve (V)boards 3D Ec 3E arc ergized th motor enerator)sets 3D 3Dh, 3ES, and 3Eh in service.Sec.s~<<"~~L'4~)~)4-~Be<(sr>s.s.7 4.Thc 250-V shutdown board 3EB battery, all three unit battcriea, a battery!charger for each battery, and associated battery boards arc OPERhBLE.Sc4>~ski((ek,oe fur C44 pg cfor OP~(pre g~q 4.a.(Delctcd)b.Once every 18 months,)the conditions under which the loss of voltage and degraded voltage relays are regni,rcd shall be simulated with an undervoltage on each shutdown board to deaoastrate that thc associated diesel generator will start.><3uStigi'agon fir Cfu~eS~SPu IST5 s BPS Unit 3 3.9/4.9W pAGE~AMENlpjp~TN 1 5 8

PERA S S.SPe.c.,g;~I'on Z.5.l 1%lOV 18 1988 10.When one 480-V shutdown board is found to be inoperable, the reactor vill be placed in HOT STANDBY CONDITION vithin 12 hours and COLD SHUTDOWH OHDITIOH within 24 hours.WStjCicah'on 4r Changing~r BAN I$75 P.g,-7 11.If on 480-V RMOV board g set s inoperab e, REA OR PO OPERATIO may co inue for perio not to exceed se en days p ovided th remain ng 80-V RMOV board sets and their associa d load remain 0 ERABLE.12 If any tvo 480-RMOV board sets ecome inop abLe, t e react r sha be pla ed in t e CO SHUTDO CONDI OH vithin 24 h urs.13.If t e r cerements or operation in the conditions specified by 3.9.B.1 through 3.9.B.12 cannot be met, an orderly shutdovn shall be initiated and the reactor shall be in the COLD SHUTDOWH CONDITION vithin 24 hours.SeC g>~>.;~Hon 4.Cg~~CS+~ON I ST~SecAen Xg BFN Unit 3'MEtf0t,1ENT M.y p g PAGE~OF 11 FEB P 7 t991 3~5~4.5.D u 1.The equipme area cooler associated vi h each RHR ump and the e uipment a ea cooler ass ciated vi each set of core spra pumps (A an C or B d D)must b OPERAB at all tim vhen the ump or pum served by hat specif cooler is c sidered t be OPERABLE'. E h equipment rea cooler is crated in c unction vith e equipment erved by that articular c oler;therefore the equipm t area coole are teste at the same fre ency as th umps vhich th serve.2.en an equipme area coo er is not OPE LE, the p p(s)served y tha cooler ust be consi ere inoperabl for Technical Specificati purposes.C Co I ky);cubi l$)Prop seA Alod Pr SR E.g.l,8 The system shall be OPERAB enever there is irra ated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWH CONDITION or as specified in pecification 3.5.E.2.OPERABILITY shall be deter-mined within 12 hours a te reactor steam pressure 4 oM reaches 150 psig from a OLD CONDITION, r a t ely PR R TO ST TUP b usi an aux iary ste sup y.~S~.~r a.PropoSed hk K 4r SR.g.g.l.HPCI Subsystem testing shall be performed as fo vs++al or Simulated Once/18 Automatic months Actuation Test b.Pump 5R~~t.g OPERA-BILITY Per Specification.O.MM c.otor per-e a ed Va ve S cif cation OP RABIL Y 1.~]5/P 3.5.I.1 Prcfo+g Qpg d.Flow Rate at Once/W no a rect r 3 v s 1 kz.op a ng pr s re qco H l~/o.BFN Unit 3 PlO 3.5/4.5-1 AMENDMENT tl0 152 ls 0 on 3 So t FEg 0 7)99t~~R z.s.l,g e.Flow Rate at Once/18~psig months Ps g The HPCI pump shall deliver.at least 5000 gpm during, each flow rate test.l.'hov l4 5R p.g.(.~Verify that Once/Wm~each valve (manual, power-operated, or automatic) in the infection flow-path that is not locked, sealed, or otherwise secured in position, is in its correc osition.A7 2~tfo+std QCTTo&3~PtChor15 G-0 H F.R If the HPCI s stem is inoperable, t e reactor may remain in ope ation f period not to exceed'ys, provided the RHR LPCI OPE LE ICS are er'.Ac z~cd iR Ic 1't If Specifications 3.5.E.1 or 3.5.E.2 are not met, the reactor vessel pressur shall be reduced to 150'<<s psig or less within 24-3g ours.L2.Co e solation Cooli t RC CS cept hat an automatic va ve ca ble of utomatic ret rn to ts ECCS osition when an ECC signal s prese t may b in a positi for a other m de of opera ion.F.Reactor Core Isolation Coolin BFN Unit 3 The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY sha Sr 8-8 l5T$p.5$1.RCIC Subsystem testing shall be performed as follows: a.Simulated Auto-Once/18 matic Actuation months Test AMENDMENT NO.I 5 2 PAG~oF (1)PRIOR TO STARTUP from a COLD CONDITION, or,<PP~o&'li'Q 2./kh'ons Algol 3.ikhon g (2)whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 485 psig, except in the COLD DOWN CONDITION or as specified in 3.5.G.2 a 3.5.G.3 below.le 5 With one of the aSove required ADS valves inoperable, provided the H CI system, the core spray system and the LPCI syst are OPERABLE, s ore e inoperable ADS valve to OPERABLE status within 14 days or be in at least a HOT SHUTDOWN CONDITION within the next 12 hours and reduce reactor s earn dome ressure to Z psig within ho s.ISb g.3L With two or more of the above required ADS valves inoperable, be in at least a HOT SHUTDOWN CONDITION within 12 hours and reduce react r steam dome pressure to g~psig within hours.IW PC pO yA QI L5 R<oooor<d pioiion p,i pico 3.0.,3)t'ai t 4iA 7J)f5~DC'3 Lo>thin l3~~in: oi oooo~n 3 i gon<s~i.i> k 1.Six valves of the Automatic Depressurisation System shall be OPERABLE: 1.Durin each o crating cycl e following tests shall be performed on the ADS: tooaL Or gg p g1~oa.A simulated automatic actuation test shall be performed PRIOR TO STARTUP a ter~ch refute.ng outa e.nua surveys anc o the bpliefgval es is over+in g 4.6..2.ppapSC pie&~Sa 3.S.l l~.gi Pcafo5cJ QcQoA F BFN Unit 3 3.5/4.5-16 hNENOMENT NO.178..;.;;tx.,--(s EC 0'7 1994 3.6.C 4.6.C Z.Anytime irradiated fuel is in the reactor vessel and reactor coolant temperature is above 2LZ'F, both the sump and air sampling systems shall be OPERABLE.From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours for the sump system or 72 hours for, thc air sampling system.The air sampling.system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance vithout providing a temporary monitor.2.With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.>+c X~SWkimh on Q~Ch~6I ggp()5 fg y q 3.6.D.3.If the condition in 1 or 2 above cannot be met, an orderly shutdovn shall be initiated aad the reactor shall be placed in thc COLD SHUTDOWN CONDITION vithin 24 hours.1.When more thaa one relief valve is kaovn to be failed, an orderly shutdovn shall bc initiated and the reactor depressurised to less than lOS psig vithin 24 hours The relief valves are not required to be OPERABLE in the COLD SHllTDOMN CQNDITIQN. 4.6.D.1.Approximately one-half of all relief valves shall be bench-checked or replaced vith a bcnchmhecked valve each operating cycle.hll 13 valves vill hav been checked or replaced upon the completion of every second cycle.<<<~r;~yon g~dtegCS 4<81-N l5T5 gag'R.g.]kr~rct jar fC."sR r.5.1.Ll ti t ermo oup e an a oust c mon tore do str am of hc ve ind cate stcam e lo ag f om the velvr 2.In accordance vith Specification 1.0.%5 c relief valve shall be manually o cned BPÃUnit 3 3.6/4.6-10 ANENMOrr NL I 88 GOOF

~Pec,'o~-S-t NY 9 l994 Whenever t core 8 ay systems, LPCI, HPCI, r RC required to be OPERAB the dis ar piping from'he ump disc rge these systems o the las bl ck valve shall filled.The s tion of the P pumps s all be aligned o the condensa storage tank, and t e press suppression amber hea tank s 1 normally b align d to se the dischar e piping f the and CS pump~e cond sate hea tank may be'us d to se e the aad CS dis harge px ing if th PSC head tank is unava able.Th pre88 e indica ors on th discha ge of the and CS umps shall i icate not less than listed b ow.1.Eve e RHRS CI and Conta nment Spray)and core spray systems, the dischar e pipin of these systems shall vened om/he h h pbjnt d wattr fl w%term ed.2.Following any period where the I or core spray sy tems have ot been equired o be OPERAB the scharge p ing of the ino erabl system 8 ll be vented f m the igh poin prior to the turn of e 8 8 tern't sly.s;/.1 The following surveillance requirements shall be adhered to assure that the discharge piping of the core spra systems, LPCI, HPCI, and RCIC are filled: Pl-75-0 Pl-75-4 Pl-74-51 Pl-74-65 48 p ig 48 ps 48 psig 48 psig 3~Whenever the HPCI R system is lined up to take suction from the condensate storage tank, the dischar e piping of the HPCI d RCI><Xsgg;mg.~Jr Chu@P~SFN ls T$p.g.p e vened from t high oint f the 8+tern and wate low observe on a monthly bas s.en the RHRS and the CS are r uired to b OPERABLE, t pre ure indicat s which monit the discha lines shall be onitored da y and he pressu e recorded.BFN Unit 3 3.5/4.5-17 NENOMENTRO, r 7 g q+Z'~QWI JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.Five current LCOs, 3.5.A, 3.5.B, 3.5.E, 3.5.G, and 3.5.H, have been combined into one proposed LCO (3.5.1).As such, the new LCO combines the three ECCS spray/injection Systems (HPCI, LPCI, and CS)into one LCO statement. The Bases continue to describe what components make up an ECCS subsystem. The new LCO statement also specifies that the six ADS valves are required.In addition, the ADS valve cycling requirements located in current Specification 4.6.D.1 are included as part of ADS operability. Thus, if an ADS valve does not cycle, the affected ECCS system is considered inoperable and the appropriate ACTION taken.A3 The Frequencies of"Once/operating cycle,""during each operating cycle," and"after each refueling outage" have been changed to"18'onths." This is considered equivalent since 18 months is the length of an operating cycle or a refueling outage cycle.The Frequencies of"Once/3 months" and"Per Specification I.O.HM" have been changed to"92 days," or"In accordance with the Inservice Testing program" as appropriate. The IST program test frequency for pumps is every 3 months and is currently defined by Specification I.O.MM.Therefore, this change is considered administrative in nature.The Frequency of"Once/month" has been changed to"31 days." 0 A4 Notes allowing actual vessel injection or ADS valve actuation to be excluded from this test (simulated automatic actuation test)have been added to proposed SR 3.5.1.9 and SR 3.5.1.10.Since the current BFN-UNITS 1, 2, 5 3 Revision 0 PAGE OF 0~I JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING I requirements state the test is"simulated" (i.e., valve actuation and vessel injection are inherently excluded), this allowance is considered administrative in nature.AS Proposed Condition H provides direction for various interrelationships between HPCI and ADS, and between LPCI and CS.The Action requires entry into LCO 3.0.3 for various combinations of inoperability which are consistent with the present required actions for the same various combinations. The actual requirements are not being changed.A6 The existing Applicability for Core Spray System (CSS)Operability (3.5.A.1), and Low Pressure Coolant Injection (LPCI)Operability (3.5.8.1), requires both systems to be Operable whenever irradiated fuel is in the vessel and prior to startup from a COLD CONDITION. The proposed change (LCO 3.5.1 Applicability) requires them to be Operable in Modes 1, 2 and 3.This change more clearly defines the conditions when CSS and LPCI are required to be Operable without changing the specific requirements which are currently located in individual specifications for each system.This change is, administrative because the same requirements for Operability currently listed in specific specifications will be labelled APPLICABILITY and apply to the entire ISTS Section 3.5.1, ECCS-Operating. The 3.5.A.2, 3.5.B.2, and 3.5.B.7 Applicabilities are only cross references and have been deleted.A7 The clarifying information contained in the"*" footnote has been moved to the proposed Bases for SR 3.5.1.2.The intent of the surveillance is to assure that the proper flow paths will exist for ECCS operation. The Bases clarifies that a valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.As such, moving this clarifying statement to the Bases is an administrative change.AS This requirement has been deleted since it only provides reference to another Specification, and does not provide any unique requirements. The format of the proposed BFN ISTS does not include providing"cross references." A9 Surveillance Requirements for HOV operability, and check valves that are required by the Inservice Testing (IST)Program, have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.BFN-UNITS 1, 2, 5 3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING A10 The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system.Since the reactor steam dome pressure must be a 920 psig to perform SR 3.5.1.7 and a 150 psig to perform SR 3.5.1.8, sufficient time is allowed after adequate pressure is achieved to perform these tests.This is clarified by a Note in both SRs that state the Surveillances are not required to be performed until 12 hours after the specified reactor steam dome pressure is reached.CTS 3.5.E.1 already contains the context of the Note for the low pressure flow rate test.This is also consistent with interpretation of the current technical specification requirement for the high pressure flow rate test which is currently not modified by a Note.All The existing Applicabilities for High Pressure Coolant Injection (HPCI)Operability (3.5.E.1)and ADS (3.5.G.1)require the systems to be Operable whenever irradiated fuel is in the vessel and reactor pressure is greater than 150 psig (105 psig for ADS), except in the COLD SHUTDOWN CONDITION. The proposed change (LCO 3.5.1 Applicability) requires HPCI and ADS to be Operable in Modes 1, 2 and 3, except when reactor steam dome pressure is<150 psig.(Reference Justification L5 for the.change in applicability from<105 psig to<150 psig for ADS.)This change more clearly defines the conditions when HPCI and ADS are required to be Operable without changing the specific requirements which are currently located in the individual specifications. This change is administrative because the same requirements for Operability currently listed in the specific specifications will be labeled APPLICABILITY and apply to the entire ISTS Section 3.5.1,,ECCS-Operating. The 3.5.E.2, 3.5.G.2, and 3.5.G.3 Applicabilities are only cross references and have been deleted.A12 A finite Completion Time has been provided to verify RCIC OPERABILITY. The new.time is immediately and is considered administrative since this is an acceptable interpretation of the time to perform the current requirement. A13 CTS 3.9.A.3.h (for Unit 1 and 2)and 3.9.A.3.g (for Unit 3)require 480 V reactor motor operated valve (RMOV)boards to be energized with motor-generator (MG)sets in service.CTS 3.9.B.13 and 14 (for Unit 1 and 2)and ll and 12 (for Unit 3)provide Required Actions for when one or any two 480-V MG board sets become inoperable. There are two 480-V AC RMOV boards that contain MG sets in their feeder lines.The 480-V AC RMOV boards provide motive power to valves associated with the LPCI mode of the RHR system.The MG sets act as electrical isolators to prevent a BFN-UNITS 1, 2, 5 3 3 Revision 0 PAGE

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING fault propagating between electrical divisions due to.an automatic transfer.Having an MG set out of service reduces the assurance that full RHR (LPCI)capacity will be available when required, therefore, the unit can only operate in this condition for 7 days.Having two MG sets out of service can considerably reduce equipment availability; therefore, the unit must be placed in Cold Shutdown within 24 hours.The'nability to provide power to the inboard injection valve and the recirculation pump discharge valve from either 4 kV board associated with an inoperable MG set would result in declaring the associated LPCI subsystems inoperable and entering the Actions required for LPCI.Since, the out of service times for LPCI and the MG sets are comparable, the deletion of the MG set actions is considered administrative. TECHNICAL CHANGE-MORE RESTRICTIVE Ml Proposed Action H requires LCO 3.0.3 be entered immediately which requires the plant to be in MODE 2 in 7 hours and MODE 3 within 13 hours when multiple ECCS subsystems are inoperable. This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.5.A.3, 3.5.B.4, 3.5.B.8, and 3.5.E.3).For CTS 3.5.G.2 it is slightly more restrictive since it requires the plant to be in MODE 2 in 7 hours where no action was required before.CTS require a shutdown to NODE 4 within 24 hours (except CTS 3.5.G.2 for ADS which also requires the plant be in NODE 3 in 12 hours)but does not stipulate how quickly MODE 3 must be reached.Reference Comment L12 which addresses the less restrictive change of being in NODE 3 in 13 hours versus 12 hours and NODE 4 (or<150 psig which is outside the applicability for ADS and HPCI)in 37 hours rather than 24 hours.Surveillance requirement SR 3.5.1.3 has been added to verify that ADS air supply header pressure is z 90 psig.This is a new Surveillance Requirement which verifies that sufficient air pressure exists in the ADS accumulators/receivers for reliable operation of ADS.Since this is a new Surveillance Requirement, it is an added restriction to plant operations. N3 With the reactor pressure<105 psig, CTS 3.5.B.2 allows the RHR System to be removed from service (except that two RHR pumps-containment cooling mode and associated heat exchangers must remain OPERABLE)for a period not to exceed 24 hours while being drained of suppression chamber quality water and filled with primary coolant quality water provided BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING that during cooldown two loops with one pump per loop or one loop with two pumps, and associated diesel generators, in the core spray system are OPERABLE.This appears to be an exception to CTS 3.5.A.2 8 3, which only allows one CSS loop (i.e., one loop with two pumps)to be inoperable for 7 days and an immediate shutdown if this cannot be met.The¹Note for 3.5.B.1 allows LPCI to be considered OPERABLE.during alignment and operation for shutdown cooling with reactor steam dome pressure<105 psig in HOT SHUTDOWN, if capable of being manually realigned and not otherwise inoperable. Proposed Specification 3.5.1 has a similar provision (Note to SR 3.5.1.2).Since the proposed Specification has no provision that would allow continued operation in MODE 3 with pressure<105 psig with two CS loops with one pump per loop OPERABLE, the proposed change is considered more restrictive. M4 An additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specifications (CTS 3.5.A.3, 3.5.B.4, 3.5.B.S, and 3.5.E.3).CTS require a shutdown to MODE 4 within 24 hours but does not stipulate how quickly MODE 3 must be reached.Reference Comment L2 which addresses the less restrictive change of being in MODE 4 (or<150 psig for HPCI and ADS)in 36 hours rather than 24 hour s.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl Not used.LA2 The details relating to system design and purpose have been relocated to the Bases.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the FSAR will be controlled by the provisions of 10 CFR 50.59.ECCS system operability determinations are described in the Bases.SR 3.5.1.1 will ensure maintenance of filled discharge piping.BFN-UNITS 1, 2,&3 Revision 0 PAGE jz JUSTIFICATION FOR CHANGES BFN ISTS 3.5-1-ECCS-OPERATING LA3 Details of the methods of performing surveillance test requirements and routine system status monitoring have been relocated to the Bases and procedures.'hanges to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA4 LA5 Any time the OPERABILITY of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. Therefore, explicit post maintenance Surveillance Requirements have been deleted from the Specifications. Also, proposed SR 3.0.1 and SR 3.0.4 require Surveillances to be current prior to declaring components operable.CTS 3.5.D/4.5.D, Equipment Area Coolers, are being relocated to plant procedures. Relocating requirements for the equipment area coolers does not preclude them from being maintained operable.They are required to be operable in order to support HPCI, RCIC, LPCI and CS system operability. If they become inoperable, the operability of the supported systems are required to be evaluated under the Safety Function Determination Program in Section 5.0 of the Technical Specifications. This change is consistent with NUREG-1433. LA6 CTS 3.5.E specifically states that HPCI Operability can be determined prior to startup by using an auxiliary steam supply in lieu of using reactor steam after reactor steam dome pressure reaches 150 psig.Details of the methods of performing this surveillance test requirement have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA7 CTS 4.5.H.l requires the discharge piping of RHR (LPCI and Containment Spray)to be vented from the high point and water level determined every month and prior to testing of these systems.The specific requirement to vent prior to testing has been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.BFN-UNITS 1, 2, 8L 3 PAGE~OF~i;; 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING"Specific" Ll The phrase"actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying the ECCS subsystems/ADS actuate on an automatic initiation signal.This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill this requirement. Operability is adequately demonstrated in either case since the ECCS subsystems/ADS itself can not discriminate between"actual" or"simulated." L2 The time to reach MODE 4, Cold Shutdown (for LPCI and CS)and<150 psig (for HPCI and ADS)has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within, the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added for LPCI, CS and HPCI (Reference Comment M4 above).These times are consistent with the BWR Standard Technical Specifications, NUREG 1433.A new Action (proposed ACTION 0)is being added to LCO 3.5.1 for the.condition of an inoperable HPCI System coincident with one inoperable low pressure ECCS injection/spray subsystem. The analysis summarized in the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996)demonstrates that adequate cooling is provided by the ADS system and the remaining operable low pressure injection/spray subsystems. However, the redundancy has been reduced such that another single failure may not maintain the ability to provide adequate core cooling.Therefore, an allowable outage time of 72 hours has been assigned to restore either the inoperable HPCI system or the inoperable low pressure injection/spray subsystem to operability. This change is consistent with NUREG-1433. L4 The allowable outage time for HPCI has been extended from 7 days to 14 days.Adequate core cooling can be provided by ADS and the low pressure ECCS subsystems. The 14 days is allowed only if all six ADS valves and the low pressure ECCS subsystems are operable.(The exception, LCO 3.5.1, Condition D, which allows operation for 72 hours with HPCI and one low pressure ECCS subsystem inoperable is addressed in Comment L3 above.)The 14 day Completion Time is based on the reliability study that evaluated the impact on ECCS availability (Memorandum from R.L.Baer (NRC)to V.Stello, Jr.(NRC),"Recommended Interim Revisions to BFN-UNITS 1, 2, 5.3 Revision 0 PAGE~GP~ 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING LCOs for ECCS Components," December 1, 1975).Factors contributing to the acceptability of allowing continued operations for 14 days with HPCI inoperable include: the similar functions of HPCI and RCIC, and that the RCIC is capable of performing the HPCI function, although at a substantially lower capacity;the continued availability of the full complement of ADS valves and the ADS System's capability in response to a small break LOCA;and, the continued availability of the full complement of low pressure ECCS subsystems which, in conjunction with ADS, are capable of responding to a small break LOCA.This change is consistent with NUREG-1433. L5~L6 The pressure at which ADS is required to be operable is increased to 150 psig to provide consistency of the operability requirements for.HPCI and RCIC.equipment. Small break loss of coolant accidents are not analyzed to occur at low pressures (i.e., between 105 and 150 psig).The ADS is required to operate to lower the pressure sufficiently so that the LPCI and CS systems can provide makeup to mitigate such accidents. Since these systems can begin to inject water into the reactor pressure vessel at pressures well above 150 psig, there is no safety significance in the ADS not being operable between 105 and 150 psig.\A new ACTION has been added (ACTION F), which allows an outage time of 72 hours when one ADS valve and a low pressure ECCS subsystem is inoperable. Currently, there is no allowed outage time when these two items are inoperable. The analysis summarized in the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996)demonstrates that adequate cooling is provided by the HPCI and the remaining operable low'pressure injection/spray system.However, the redundancy has been reduced such that another single failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Therefore, an allowable outage time of 72 hours has been assigned to restore either the inoperable ADS.valve or the inoperable low pressure injection/spray system.This change is consistent with NUREG-1433. L7 t out of service as BFN-UNITS 1, 2, 8.3 Revision 0 Current Technical Specifications only allow one LPCI pump to be inoperable. Proposed ACTION A allows two LPCI pumps, one per loop or two in one loop, to be inoperable for seven days.The BASES for ISTS 3.5.1 Required Action A.l state that the 7 day allowed outage time is justified because in this condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA.This justification is applicable for the LPCI function of RHR with one or two RHR (LPCI)pumps demonstrated by previous LOCA analyses performed for

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING BFN as well as the current SAFER/GESTR-LOCA analysis (NEDC-32484P, February 1996).Following postulated single failures, adequate core cooling can be provided by one loop of Core Spray (2 pumps)and two RHR (LPCI)pumps (either two pumps in one loop or one pump in two loops)in conjunction with HPCI and ADS.Therefore, this less restrictive change is acceptable based on the plant specific LOCA analysis perfqrmed for BFN.L8~L9 This change proposes to add a Note to current Surveillance Requirement 4.6.D.4 (proposed Surveillance Requirement 3.5.1,12)which states,"Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test." This change allows the Applicability of the Specification to be entered for 12 hours without performing the Surveillance Requirement. This allows for sufficient conditions to exist and allow the plant to stabilize within these conditions prior to performing the Surveillance. The normal outcome of the performance of a Surveillance is the successful completion which proves Operability. This change represents a relaxation over existing requirements. This change is consistent with NUREG-1433. Existing Surveillance Requirement 4.5.E.l.d requires verification that HPCI is capable of delivering at least 5000 gpm at normal reactor vessel operating pressure.The proposed surveillance, SR 3.5.1.7, requires verification of a minimum 5000 gpm HPCI flow rate with reactor pressure e 920 psig and<1010 psig.The HPCI performance test at high pressure is the second part of a two part test that verifies HPCI pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the HPCI pump is expected to perform.Performance of the HPCI test at both ends of the expected operating pressure range confirms that the HPCI pump and turbine are functioning in accordance with design specifications. The ability of the HPCI pump to perform at normal reactor vessel operating pressure has already been demonstrated. A small decrease in the pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. BFN-UNITS 1, 2, tIL 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING L10 Existing Surveillance Requirement 4.5.C.l.e requires verification that HPCI is capable of delivering at least 5000 gpm"at 150 psig reactor steam pressure." The proposed surveillance, SR 3.5.1.9, requires verification of a minimum 5000 gpm HPCI flow rate with reactor pressure at a 165 psig.This change is less restrictive because it could allow reactor operation at pressures up to 165 psig prior to performing the surveillance. Performance of HPCI pump testing draws steam from the reactor and could affect reactor pressure significantly. Therefore, HPCI pump testing must be performed when the Electro-Hydraulic Control (EHC)System for the main turbine is available and capable of regulating reactor pressure.Operating experience has demonstrated that reactor pressures as high as 165 psig may be required before the EHC system is capable of maintaining stable pressure during the performance of the HPCI test.The HPCI performance test at low pressure is the first part of a two part test that verifies HPCI pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the HPCI pump is expected to perform.Performance of the HPCI test at both ends of the expected operating pressure range confirms that the HPCI pump and turbine are functioning In accordance with design specifications. The ability of the HPCI pump to perform at the lowest required pressure of 150 psig has already been demonstrated. A small increase in the pressure at which the performance to design specifications is verified will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. Ll1 CTS 3.5.E.1 requires HPCI operability to be determined within 12 hours after reactor steam dome pressure reaches 150 psig from a COLD CONDITION. The proposed Note to SR 3.5.1.7 and 3.5.1.8 allows X2 hours to perform the test after reactor steam dome pressure and flow are adequate.This is based on the need to reach conditions appropriate for testing.The existing allowance to reach a given pressure only partially addresses the issue.This pressure can be attained, and with little or no steam flow, conditions would not be adequate to perform the test-potentially resulting in an undesired reactor depressurization. The proposed change recognizes the necessary conditions of steam flow and minimum pressure as well as a maximum pressure limitation and provides consistency of presentation of these conditions. The point in time during startup that testing would begin remains unchanged. The change simply changes when the 12 hour clock for performing the test 10 Revision 0 PAGED()p~ JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1>>ECCS-OPERATING must begin and permits testing to be completed in a reasonable period of time.L12 Proposed Condition H provides direction for various interrelationships between HPCI and ADS, and Between LPCI and CS.The Action requires entry into LCO 3.0.3 for various combinations of inoperability which are consistent with the present required actions for the same various combinations (CTS 3.5.A.3, 3.5.B.4, 3.5.B.8, and 3.5.E.3).However, the time to reach MODE 4, Cold Shutdown (for LPCI and CS)and<150 psig (for HPCI)has been extended from 24 hours to 37 hours and to reach MODE 3, Hot Shutdown (for ADS only)has been extended from 12 hours to 13 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 2 in 7 hours and MODE 3 (Hot Shutdown)within 13 hours has been added (Reference Comment Ml above).These times are consistent with the BMR Standard Technical Specifications, NUREG 1433.~L13 An alternate verification to ensure the LPCI cross tie between loops is isolated has been added for Unit 3.The addition of an alternate method of satisfying the surveillance requirement is considered less restrictive. Currently, the method used for all three units is to verify the LPCI cross tie is closed and power is removed from the valve operator.Unit 3 has a manual shutoff valve install between the cross tie for Loop I and Loop II.This verification ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other subsystem. Since the manual shutoff valve serves the same function as the power operated valve, the proposed change is considered acceptable. BFN-UNITS 1, 2,&3 Revision 0 PAGE~(PP (P S JUSTIFICATION FOR CHANGES BFN ISTS 3.5.1-ECCS-OPERATING RELOCATED SPECIFICATIONS Rl Browns Ferry Nuclear Plant consists of three units.The pump suction and heat exchanger discharge lines of one loop of RHR in Unit 1 (Loop II)are cross-connected to the pump suction and heat exchanger of Unit 2.Unit 2 and 3 systems are cross-connected in a similar manner.Technical Specification requirements related to RHR cross-tie capability between units have been deleted.The standby coolant supply connection and RHR crossties are provided to maintain long-term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR System associated with a given unit.They provide added long-term redundancy to the other ECC Systems and are designed to accommodate certain situations which, although unlikely to occur, could jeopardize the functioning of these systems.Neither the RHR cross-tie nor the standby coolant supply capability is assumed to function for mitigation of any transient or accident analyzed in the FSAR.Therefore, the operability requirements and surveillances associated with the cross-connection capability have been relocated to the Technical Requirements Manual (TRM).Changes to the TRM will be controlled in accordance with 10 CFR 50.59.0 BFN-UNITS 1, 2, 5 3 12 Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP ,4 AUG 02 1989 3.5.A Co 4.5.A Co S a S st SS 5'mS~C'~+on @c Cho,+45 Q(Qpg)5+5 p 4.5.A.l.d (Cont'd)105 psi dif'fcrential pressure betveen thc reactor vessel and the primary containment. e.Check Valve Per Specification 1.0.MM 2.If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days providing all active components in the other CSS loop and the RHR system (LPCI mode)and the diesel generators are OPERABLE.Once/Month f.Verify that each valve (manual, povcr-operated, or automatic) in the injection flovpath that is not locked, scaled, or other-visc sccurcd in position, is in its correct+position.3~If Specification 3.S.A.1 or Specification 3.5.A.2 cannot bc met, the reactor shall be placed in the COLD SHUTDOWH COHDITIOH vithin 24 hours.2.Ho additional surveillance is required.4~Ql;caL: l.g LCo Ze 5.g When thc reactor vessel pressure is atmospheric and irradiated fuel is in the eactor vessel at least one core spray loop vith one OPERABLE um associated eccl generator shall be PERABLE except vit the reactor vcsscl head removed as specified in 3.5.A.5 r TO STARTUP as spccificd in 3.5.A.1.Except that an automatic valve capable of automati return to its ECCS positi vhcn an ECCS signal is present may be in a position for another mode of operation. Wc'ssFi(aH ~*a Clonic<4(BC'Sos l.S.J S~5<S4$e(a~~C~g~BPH 15'f5 34a2.BFH Unit 1 3'/4 5-2 FAGE~oF-7-hMENDMENT NO.16 9 4l Al 5fec;0;c)hoz r.s',z QEI: 15 f988 LCo~l cab'.Lsd Mhen irradiated fuel is in the reactor vessel and the zeac'tor vessel head is removed, core spray is not required to be OPERhBLE provided the cavity is flooded, the fuel pool gates are open and the fuel pool vater level is maintained above the lov level alarm point and rov one V ump ass ciate valv su ply thc andb coo ant s ply are OPERhBLE Profosc'L/tCT'I oW5'Its PssW SR 3 S'.Pisl'sos>~IF Z.5.z.S'r GSS Mhcn vork is in progress vhich has the potent al to drai the vessel, ual nitia on apabilit of ei er 1 SS L p or 1 pum vith he ca bility o injec ng va into he re cl assoc ate ese generator(s) are required.Me SusHk~m~on 0 r C~C 4o s~>mrs z,~.~BFK Unit 1 3.5/4.5-3 AMENDMENT 10.y 6 g kl 8.9.APQ'cab Jig Mo 3.5o2.Nag For SR xs;zA If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be~initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION ithin 24 hours.When the reactor vessel pressure is atmospheric and irradiated fuel is in the eactor vessel, at least one RHR loop with two pumps or two loops with one pump per loo shall be OPERABLE.c pumps soc ate mesc generators ust also be OPERABLE.prcssure coo an xngcction (LPCI)may be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 8.No additional surveillance required.<r<rus+Amaon P r r h~W B<H t S'TS'R.s.l~R 9.When the reactor vessel pressure is atmospheric, the RHR pumps that are required to e OPERABLE shall be Z.demonstrated to be OPERABLE per Specification 1.0.MN.St'~3uSA/'cocoon Pg,r Cha~t Ac BPN ISIS g.g.a LCu Hept:ab;l+ If thc conditions of Specification 3.5.A.5 are met, LPCI and containment cooling arc not re uired.nce BFN Unit 1 When there is irradiated fuel in the reactor and thc reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The RHR pumps on the adjacent units which supply cross-connect capability shall be.demonstrated to be OPERABLE per Specification 1.0.MN when the cross-connect capability is re uired pre S~6C:~o ger Cha~S 4 BAN isTs z.s.('ANENOMENT NO.2 P 4 PAGE OF 4 -~i~e~eiiL%9KI~)i~i4ar aa4I~ri-4~ilier rt~f'lL I O'P PL J'O'I't O'I 9'1~l.~.~-~'~~~~.~II~~'~~~~II~~II~~~~~~~~I~~~~~~II~~~~~II~~~~~~~~~II~~~~~~~'I~~~~'~~~~I~~~'~~~~I~'~~~II~II~Cb~I~~~I.'t'I~~~~~~~~~~'~'~'~~~~~~~II~'~~~~~~:I'I~~~~~i~'~II~~~~~~~~~~~~~~II~~~~~~~'~II~~~~II'~~~~~II~II~~I~'~~~~~~~'~'I~II~~~~~~~

~<Wkstigic~C~g hsc B~I 5 3,ge R<<(+2 LIMITIHG COHDITIOHS FOR OPERATIOH sFecl+;cg,go& 7 s VEILLAHCE REQUIREMEHTS 3.7 4.7 cab Applies to the operating status of the primary and secondary containmcnt systems.Applies to thc primary and secondary containment integrity. OOQ~LvV To assure the integrity of the primary and secondary containment systems.To verify the integrity of the primary and secondary containment. A.C a S<3.S.~.Ag~t;~4 Leo@, g,g At any time that thc irradiated fuel is in th reactor vessel, an the nuc car s em s pressurized ov tmos hcric ressure or work is being done v has the potential to drain the vessel, thc pressure su 1 vatcr level d tern eratur s a e maintained vithin the folloving limits.a.Minimum water level~-6.25" (differential pressure control>0 paid)-7.25" (0 paid differen-tial pressure control)S'C Z,S,2>I ai Thc suppression chamber vater level be checked once cr enever heat s added to the suppression pool by testing of thc ECCS or relief valves the pool temperature shall be continually monitored and shall'be observed and logged every 5 minutes until the heat addition is terminated. Max~1N b imum vater level~BFH Unit 1 3.7/4.7-1 PAGE i 3..C.S utdo 4.9~G~0 S do Whenever the reactor is'n COLD SHUTDOWH COHDITIOH vith irradiated fuel in the reactor, the availability of electric power shall be as specified in Section.3.9.A except as specified herein.l.At least tvo units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be OPERABLE.l.Ho additional surveillance is required.See AsÃkcah~~N CQ yy 8~<lS'75 Sec+io~7,f 2.An additional source of pover energized and capable of supplying power to the units 1 and 2 shutdovn boards consisting of at least one of the following: a.One of the offsite pover sources specified in 3.9.A.1.c. b.A third OPERABLE diesel generator. 3.At least one 480-V shutdown board for each unit must be OPERABLE.4.One 480-V RMOV boar mg t is uire for ach boar (1D o 1E)ired t suppo t oper tion o the syst in ac ordanc vit 3.5.B.BFH Unit 1 3.9/4.9-15 AMENDMENT tttO, 2 0 3;-"..~~r'F

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE~OF~

Cr.X'C.4;o 3.K.Q AUG OR 1998 3 5.h 4~5.h S SS Scc.awc$;P,~g,.8~~+~Sr's~8 s/4-5-h l.d (Cont'd)105 psi differential pressure betveen the reactor vessel and the primary containment. e.Check Valve Per Specification 1.0.MM 2.If one CSS loop is inoperable, the reactor may remain in operation for a period not to exceed 7 days providing all active components in the other CSS loop and the RHR system (LPCI mode)and the diesel generators are OPERhBLE.Once/Month f.Verify that each valve (manual, pover-operated, or automatic) in the~ection flovpath that is not locked, sealed'r other-vise secured in position, is in its correct+position.30 If Specification 3.5.h.l or Specification 3.5.1.2 cannot be met, the reactor shall be placed in the COLD SHUTDOWN COHDITIO hours.2.Ho additional surveillance is required.When the reactor vessel ressure is atmospheric and irradiated fuel is in the eactor vessel at least one core spray loop vith one OPERABLE pump assoc ated esel generator shall be PERhBLE except vith the reactor vessel head removed as s ecified in 3.5.h.5 r PRIOR TO as specified in 3.5.h.l.Except that an automatic valve capable of automatic return to its ECCS position" vhen an ECCS signal is present may be in a position for another mode of operation. c'st t 3 sagk i(i cd i~4~C~grg W Bf~1sT<3>I~cc ZNsJAi~,f'>a~ g~Ck~ggf+~~FIJ Isis z.g.2 BPH Unit 2 3.5/4'-2 hMENWENNO. 16 9 8 5 ~Pter fi~fio~3.5.~DEC 15 l988 Lco 3.S Z.When irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required to be OPERABLE provided the cavity is flooded, the fuel pool gates are open and the fuel pool vater level is maintained above the lov level alarm point and r v e one W puhy and ssociated alves g suppl the st dby coolant supply are OPERABLE.P~opocM Sg 8.<.B.A-'Popo~S~8.S:2.~CSS p<<posW ACnoaS.*When vork is in progress vhich ha the potenti l to drai the vess l, manual i tiation capab ity of eith l CSS Loop or RHR pump, ith the capabili of injecti te 411 the assoc ate ese enerator(s) are re uired s,~z~~$;4Lc 4'~4r 0~~~3.F.2-BFH Unit 2 3.5/4.5-3 AMENDMENT g6.~g 8 PAGE~i-iF~ S ent'nmen t 8.If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hour 8.No additional surveillance required.><~~s4i4c f~4, C~rri.4r Bf'N I s~g 9.4t't~W J.k>LCo 3.5.Q en the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.e pumps associated diesel generators must also RABLE Low pressure coolant injection (LPCI)may be, considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. 9.When the reactor vessel pressure is atmospheric that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.HM.CQ O~~pl;~'.);t ~t e conditions o Specification 3.5.A.5 are met, LPCI and containment cooling re not re 404pk4%84~BFN Unit 2 When t ere is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours'~)3.5/4.5-7 The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.Sec.QNgf<f<~$ ~Qi (~~~~4 g~><sees B,g (AMENDMENT RD.2 2 3

~pc<Fice y i'o~Z.g~AUG 02 1989 IREMEHTS Whenever thc core spray systems, LPCI, HPCI, or RCIC are requi.red to be OPEBhBLE, the discharge piping from the pump discharge of these systems to thc last block valve'hall be filled.Thc suction of thc RCIC and HPCI pumps shall be aligned to the condensate storage tank, and the pressure suppression chamber head tank shall normally bc aligned to serve the discharge piping of the RHR and CS pumps.The condensatc head tank may be used to scrvc thc RHR and CS discharge piping if the PSC head tank is unavailable. The prcssure'ndicators on the discharge of the RHR and CS pumps shall indicate not less than listed belov.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 psig Pl-74>>65 48 psig Sec a4c44i~)o Q, PL QCt'r 85~Isrs 35/~353 The folloving surveillance requirements shall be adhered to assure that thc discharge'iping of the"core spray systems, LPCI , an RCI are filled.1.Every month an pr or toi e t t ng o the S (LPCI and Cont ent Spray)and co spra system the discharge p ping of these systems s a 3 e~te re e pqint and visitor flo~etermi ed 2~Folloving any per o v ere e,LPCI or co e spray systems ha e not been r ired~be OPE, the dis rge ping of th inoperable system shall be vent+from the high point prior to the return of thc system to 3.Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage tank, the discharge piping, of thc HPCI and RCIC shall be vented from the high point of the system and vater flov observed on a monthly basis.4.en the RHRS e SS are rc ired to e OPE LE, he pres re indi ators ich~p,)monito the di charge nes shall be onitored daily and the prcssure recorded.BFH Unit 2 3.5/4.5-17 AMStNENT Rt3.I 6 g PAGE

sec wgfgq,s:~P C4o~y gr gru I JMX S.g.2./f-~5 ceo ficRli~3.5,Q LIMZTIHG COHDITIOKS FOR OPE1RTI05 SURVEILLhHCE REQUIREMENTS ~7 4~7 C Applies to the operating status of the primary and secondary containment systems.Applies to the primary and secondary containment integrity. ~O~JJv To assure thc integrity of thc primary and secondary containment systems.Qhiaafze To verify the integrity of the primary and secondary containment. 1.At amr time that the irradiated fuel is in the reactor vessel and the nuclear system is pressurized above atmospheric prcssure or wor s e done whi has the potential to drain the vessel,'Ke prcssure suppression ool wats lcvc temperature ma nta ed within the following limits.a.Minianun water level~-6.25" (differential pressure control>0 paid',-7.25" (0 psid differen-tial prcssure control.SR as.z I a.The suppZession chamber water level b checked once er enever heat is added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall bc observed and logged every 5 minutes until the heat addition is terminated. b.Maximum water level~BFH Unit 2 3.7/4.7-1 OF V +CcifiC%4iue Z.S.Q VAN 0 8$99$3.9.C.0 o C d udo 4.AC 0 Co d S utdo whenever the reactor is in COLD SHUTDOWN COHDITIOH vith irradiated fuel in the reactor, the availability of electric pover shall be as specified in Section 3.9.A except as specified herein.~~ao aoclltlonsl surveillance is required.l.At least tvo Units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be OPERABLE.2.An additional source of pover energized and capable of supplying pover to the Units 1 and 2 shutdovn boards consisting of at least one of the folloving: Sea 3453 fscR'4d~Vol C lg-Jp!B~+I~~Seel~~9.f'.One of the offsite pover sources spec'fied in 3.9.A.l.c. b.A third OPERABLE diesel generator 3.At least one 480-V shutdown board for each unit must be OPERABLE.4.One 480-V RMOV board set required or each OV bo d (2D or require to supp t operatio of the RHR system accordance vith 3.5.B.9.BFH Unit 2 3.9/4.9-15 ANENOMENT RO.1 8 6 PAGE~OF~ 4 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP At deci gi earn AUB 02 lggg 4.5.k 4.5.k.l.d (Cont'd)GV&hkiWA on Ar C~(y g@g/pe g5yg 3.5.i 105 psi differential pressure between the reactor vessel and the yrimary contaimaent ~e.Testable Per Check Valve Specification Z.O.M 2.If one CSS loop is inoperable, thc reactor may remain in operatian for a period not to exceed 7 days providing all active components in thc other CSS loop and the RHR system (LPCI mode)and the dicscl generators are OPERkSLE.f.Verify that Onc each valve (manual y paver oyerated, or automatic) in the infection flowpath that is not locked, sealed, or othcr-vise secured in positiang is in its correct+position.3.If Specification 3.5.h.l or Syecificatian 3.5.k.2 cannot be met, the reactor shall be placed in the COLD SHOTDOWÃCOHDITIOE vithin 24 hours.2.No additional surveillance is re~ired.4.When the reactor vessel yressure is atmospheric and irradiated fuel is in the reactor vessel at least one (core syray laop vith one Z.S~CO umy associate esel generator shall b OPERABIZ exccyt vith the reactor vessel head rcmovcd as s ecificd in 3 AS.A.5 r RIOR as s ecificd in 3.5.k.l.Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may bc in a position for another mode of apcration. BFH Unit 3 5i-<~u&llca,giin Qp I~~A~xs 55 3.5;I 3.5/4'-2 See Z~sk:t';i,g~ 4 CIi-)~g~V Isis g.~.<PAGE~OF~NENONKNTNO. 1 O 0 Cl, Sfec.>gcW~ 3.5.2;DEC 15 1988*s.LGo gg.2 4'l'mb: lifp When irradiated fuel is in thc reactor vessel and the reactor vessel head ia removed, core spray is not required to be OPERABLE provided thc cavity is flooded, thc fuel pool gates are open and the fuel pool vater level is maintained above the lov evel alarm point ro ne p assoc atcd val cs s plying e stan y coo ant sup y are OPE LE Pw os'~s Sg, 3.g.g,g~cs~~oSect AC7 gong*When vork is in p gress vhich as thc po ential drain the ssel, man al init tion c ability o either CSS Loo or 1 pump, vi h the capa ility of a/ecting va cr into he reacto vessel c e generator(s) are required.Se<3'~f;~go~Q~gA BFH Unit 3 3.5/4.5-3 AMENOMENT N5.g P 2 Cl 8.9.LCn 3.sR X5 2,g If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop hall be OPERABLE.e pump assoc@a e xesel gener or must also be OPERABLE Low pressure coolant njection (LPCI)may be considered OPERABLE during alignment iand operation for shutdown cooling, if capable of being manually realigned and not otherwise inoperable. Sc<5<sWVi ca,Ao n&<add tsrs~.s.~SR Z.5,2,S When the reactor vessel pressure is atmospheric, the RHR pumps that are required to e OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.~c 3'u5~'eaHo~ g I C4~b~e 1STS r,8.z 8.No additional surveillance required.~o 10 A(piiu b:[Q If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling~Le not required.e'FN Unit 3 en there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.(Note: Becaus~cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)3.5/4.5-7 11.The B and D RHR pumps on'nit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when, the cross-connect capability is required.e<gu~'Fi'WHon 4r~p&4/PE 15'f5 AMENDMENT No.X 77

'c NN 1 S 1994 Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled-5g X e fol owing surveillance requirements shall be adhered to assure that the discharge piping of the core s ray systems, LPCI, HPCI, an CIC are fille The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.Eve month and prio to t esting o the RS (L I and Con ament Spra and c re s ray s ems the dischar e piping of these systems shal e v e rom e xgh po t and wa flow dete~ned.2.o owing any period where the LPCI or core spray syst ve not been req red to OP LE the disch e i x 2 g P P g of th noperable sys shall be vente rom the high oint prior to the turn of the system to;service. Pl-75-20 Pl-75-48 Pl-74-51 PI-74-65 48 psig 48 psig 48 psig 48 psig 3.Whenever the HPCI or RCIC system is lined up to take suction from the condensate 'torage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis." 4.en the RS and he CSS are r uired to be OP BLE, the pre ure in cators ich moni o the d charge ines shall e onito ed dai and the pr sure rec rded.BPÃUnit 3 3.5/4.5-17 NENDMHfT i(0.1 78 PAGE 5 QF~

Skc>g~LDGTZBC CONDITIO?N ZOR OPERLTIOS SURVEILLAECE REQUIREHEHTS 3.7 4.7 kppliea to the operating'status of the priaary and secondary containNent ayateaa.hypliea to the primary and secondary contahunent integrity. To assure the integrity of the priaary and secondary contahaent ayateaa.To veri~the integrity of the priaLary and secondary cont ainccnt~gdd walib&~35'~l.Lt any togae that the irradiated fuel ia in the reactor vessel, e xmc e s ea yressurixed shore ataoapheric yreaaure or r e one~the yotential to drain the vessel, the pressure suppression pool water level tcRpera aa ta within the'following liaita.a.Madam water lerel~-6.25" (differential pressure control>0 yaid)-7.25" (0 ysid differen-tial yreaaure control)GAS.S.z.1 a.The auyyreaaion chaaber water level be checked once per glv$enerer heat a added to the suppression pool by testing, of the ECCS or relief valves the yool temperature shall be conthmally acmitored and shall be obaerred and logged every 5 minutes until the heat addition is terainated. .Maxima water level a See Q~g f'Non Q Chang)~~~'~<>34"Z.t+a BF5 Unit 3 3.7/4.7-1 PAGE~OF~ Cl 3.9.C.0~CO DIXIE S 0 4.9.C 0~CO)~~0 0 S DOWN Whenever the reactor is in the COLD SHUTDOWH COHDITIOH vith irradiated fuel in the reactor,.the availability of electric pover shall be as specified in Section 3.9.A except as specified herein.l.At least tvo Unit 3 diesel generators and their associated 4-kV shutdovn boards shall be OPERABLE.1.Ho additional surveillance is required.sc'c'5%f.'ca5~n far c~~+~><<JSVS 5 Ao q,~2.An additional source of pover energized and capable of supplying pover to the Unit 3 shutdown boards consisting of at least one of the folloving: a.One of the offsite pover sources specified in 3.9.A.l.c. b.A third OPERABLE diesel generator. 3.At least one Unit 3 480-V shutdown board must be OPERABLE.4.One 480-RNOV b ard motor nerator (mg)se is re uired fo each OV board (3D r 3E)r uired o suppo t operat on of e RHR system n accor nce v h 3.5.B.9~BFH Unit 3 3.9/4.9-14 AMENDMEQ S~g~z~8 s.VI 13USTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN ADMINISTRATIVE CHANGES A1'eformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.Surveillance Requirements for MOV operability that are required by the Inservice Testing (IST)Program have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.A3 CTS 3.9.C.4 requires one 480 V reactor motor operated valve (RMOV)board motor-generator (MG)set for each RMOV board required to support the RHR System in accordance with CTS 3.5.B.9.The 480-V AC RMOV boards provide motive power to valves associated with the LPCI mode of the RHR system.The MG sets act as electrical isolators to prevent a fault propagating between electrical divisions due to an automatic transfer.The inability to provide power to the inboard injection valve and the recirculation pump discharge valve from either 4 kV board associated with an inoperable MG set would result in declaring the associated LPCI subsystems inoperable and entering the Actions required for LPCI.Therefore, the deletion of the operability requirement associated with the MG sets in CTS 3.9.C.4 is considered administrative. BFN-UNITS 1, 2, 5 3 Revision 0 PAGE/

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN TECHNICAL CHANGE-MORE RESTRICTIVE Ml Proposed ACTIONS A, B, C and 0 have been added to provide required actions be taken when LCO requirements can not be met.CTS 3.5.A.4 and 3.5.B.9 provide minimum requirements for ECCS subsystems when in MODE 4 and 5 (except with the spent fuel pool gates.removed and water level a the low level alarm setpoint of the spent fuel pool)but no action if these requirements are not met.Therefore, technical specifications are violated when these requirements can not be met and the default to TS 1.0.C.1 requires no action since the plant is already in Cold Shutdown.While from a compliance standpoint the proposed ACTIONS are less restrictive, from an operational perspective they are more restrictive since actions are required w'here there were none before.Proposed ACTION A allows 4 hours to restore a subsystem when only one of the required subsystems is inoperable and then proposed ACTION B requires action be initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs)immediately. The 4 hour Completion Time is considered acceptable based on engineering judgment that considers the remaining available subsystem and the low probability of a vessel draindown event during this period.With no required ECCS injection spray subsystems inoperable, proposed ACTION C requires action to be initiated immediately to suspend OPDRVs and at least one required subsystem be restored to OPERABLE status within 4 hours.If one subsystem can not be restored within four hours then Proposed ACTION D requires action be initiated immediately to restore secondary containment to OPERABLE status, to restore two standby gas treatment systems to OPERABLE status, and to restore isolation capability in each required secondary containment penetration flow path not isolated.These actions must be immediately initiated to minimize the probability of a vessel draindown and the subsequent potential for fission product release.0 M2 Proposed SR 3.5.2.1 has been added.SR 3.5.2.1 requires the suppression pool water be verified~a minimum level every 12 hours.CTS 3.7.A.1 (8 4.7.A.l.a)requires the suppression pool be verified e-6.25" with no differential pressure control once per day at any time irradiated fuel is in the reactor vessel, and the nuclear system is pressurized or work is being done which has the potential to drain the vessel.Therefore, proposed SR 3.5.2.1 is more restrictive since the frequency of performance has been increased from once per 24 hours to once per 12 hours.In addition, CTS only requires performance during atmospheric conditions when work is being done that has the potential to drain the vessel.Therefore, the proposed SR is more restrictive since it BFN-UNITS 1, 2, 5 3 Revision 0 PAULO';" 6 Cl JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN requires performance during MODES 4, and 5, except with the spent fuel storage pool gates removed and water greater than or equal to minimum level over the top of the reactor pressure vessel flange.The CTS requirement to check the maximum level during OPDRVs has not been included since Specification

3.5.2 concerns

the ability to maintain reactor water level using the suppression pool as a source of water.However, this level check is required for proposed Specifications 3.6.2.1 and 3.6.2.2 as it relates to Containment Systems.M3 Proposed SR 3.5.2.4, which requires a verification every 31 days that ECCS injection/spray valves are in their correct position, has been added.This provides assurance that the proper flow paths will exist for ECCS operation. This is more restrictive since BFN currently only requires this check during MODES 1, 2 and 3.M4 An SR has been added to require a system flow rate test for the Core Spray System during atmospheric conditions. While CTS (4.5.B.9)requires flow rate testing of the RHR pumps during atmospheric conditions as well as during MODES 1, 2, and 3, it only requires CSS flow rate testing during MODES 1, 2, and 3.The addition of this requirement is more restrictive. TECHNICAL CHANGE-LESS RESTRICTIVE"Generic" LA1 CTS 4.5.H.1 requires the discharge p'iping of RHR (LPCI and Containment Spray)to be vented from the high point and water level determined every month and prior to testing of these systems.The specific requirement to vent prior to testing has been relocated to procedures. Changes to the procedures will be controlled by the licensee controlled programs.LA2 Any time the OPERABILITY of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. Therefore, explicit post maintenance Surveillance Requirements have been deleted from the Specifications. Also, proposed SR 3.0.1 and SR 3.0.4 require Surveillances to be current prior to declaring components operable.PAGE~OF BFN-UNITS 1, 2,&3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN LA3 Details of the methods of performing surveillance test requirements and routine system status monitoring have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll CTS 3.5.A.5 requires manual initiation capability of either 1 CSS Loop or 1 RHR pump with capability of injecting water into the reactor vessel when work is in progress which has the potential to drain the vessel.The proposed Specification would not require the CSS or RHR (LPCI and containment cooling mode)system to be operable since LCO 3.5.2 applicability does not apply when the fuel pool gates are open and the fuel pool water level is maintained above the low level alarm setpoint.Therefore, the deletion of this requirement is considered less restrictive. The deletion is acceptable since the coolant inventory represented by this water level is sufficient to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. L2 The proposed LCO for ECCS-Shutdown is less restrictive since it only requires two low pressure ECCS subsystems to be OPERABLE.This can be fulfilled with any combination of RHR and CS subsystems. That is, two CS subsystems (a CS subsystem for Specification

3.5.2 consists

of at least one pump in one loop), two RHR subsystems (RHR subsystem for Specification

3.5.2 consists

of one pump in one loop), or one RHR subsystem and one CS subsystem OPERABLE.CTS 3.5.B.9 requires one RHR loop with two pumps or two RHR loops with one pumps per loop to be.OPERABLE.CTS 3.5.B.4 requires one CS loop with one pump per loop to be OPERABLE.Per CTS 3.5.A Bases the minimum requirement at atmospheric pressure is for one supply of makeup water to the core.Therefore, requiring two RHR pumps and one CS pump to be OPERABLE provides excess redundancy. In addition, since only one supply of makeup water is required, sufficient makeup water can be provided by two CS subsystems, two RHR subsystems, or one CS and one RHR subsystem. As such, the proposed Specification ensures redundancy by requiring any two low pressure ECCS subsystems to be OPERABLE.BFN-UNITS 1, 2,&3 Revision 0 JUSTIFICATION FOR CHANGES BFN ISTS 3.5.2-ECCS-SHUTDOWN RELOCATED SPECI F I CAT IONS Rl Browns Ferry Nuclear Plant consists of three units.The pump suction and heat exchanger discharge lines of one loop of RHR in Unit I (Loop II)are cross-connected to the pump suction and heat exchanger of Unit 2.Unit 2 and 3 systems are cross-connected in a similar mariner.Technical Specification requirements related to RHR cross-tie capability between units have been deleted.The standby coolant supply connection and RHR crossties are provided to maintain long-term'reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR System associated with a given unit.They provide added long-term redundancy to the other ECC Systems and are designed to accommodate certain situations which, although unlikely to occur, could jeopardize the functioning of these systems.Neither the RHR cross-tie nor the standby coolant supply capability is assumed to function for mitigation of any transient or accident analyzed in the FSAR.Therefore, the operability requirements and surveillances associated with the cross-connection capability have been relocated to the Technical Requirements Manual (TRM).Relocation to the TRM is in accordance with the"Application of Selection Criteria to BFN TS" and the NRC Final Policy Statement on Technical Specification Improvements. Refer to the application document discussion for additional information. BFN-UNITS I, 2,&3 Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF SfCC<CiCOQO m 3 i5 i3 FEB 0 7 199$3.5.E S st es ure Coo a t ectio C S 4.5.E essu e Coo a t S ste HPC S ect o 4.5.E.1 (Cont'd)Sce'uste C>>capon go>>ages Q>>gFv4 l5TS 3,5,(e.Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.f.Verify that Once/Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct*position.2.If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.*Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. eactor Co LCO P.S,>1.The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 5'g3,5g,g A'3 P~ppQ Vcr sR 3.5.3~RCIC Subsystem testing"hall be performed as follows: pc~i b C J a.Simulated Auto-Once/18 matic Actuati'on .-..oaths Test BFN Unit 1 3.5/4.5-14 AMENDMENT Ho.g 8 O=;:r'~P QF

5 ci+icgQoq Z 5'.3 NOV 24 1989 e determined vithin 12 hours after reactor steam pressure (oped reaches 150 psig from a COLD"~~~~COHDITIOH em veiny PQOQ T ST TU by ing Ran a~lory ste pl SP3.5.3.3 b.Pump OPERABILITY Per Specifi-cation 1.0.MM S R3.S,p,p PAyc se4 blue.Ai sRRs;p, c.M tor-0 era d er Va e eci PE BILI cat on 1.0.MM 9z~~T>d.Flov Rate at Once/N rma re ctor s ve sel pe ating pre ure Olo P 2~Qc7 le A IQ La 3~If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed W days if the HPCIS s OPERABLE during such time~ygi.,Q If Specifications .5.F.1 or 3.5.F.2 are not met, an~~~k.and the reactor shall be depressurized to less than 150 psig within 24'ours.ore~~~SRP.S>.q Se 3.S,q,p SR g.g,p g SR Z.S.'3.2 Be~Aodc 3 in IWhts Once/18 months f.Verify that Once each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc osition.A4, Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mod of o eration.e.Flov Rate at psig~ILS The RCIC pump shall deliver at least 600 gpm during each flov test.rlkqs BFH Unit 1 3.5/4.5-15 AMENOMENT NO.I 7 3 PAGE 8 >f'eciWi an 3,5,3 m 19 1994-Whee>~he core s ra s stems HPCI or ZC.ar required to OPERAB , t dis arge pipin from the pum disc rge the systems to e la t gg(b ck va ve shall be f lie Hm.5'g 8'.C.3.J The following surveillance requiremeats shall be adhered to assure that the discharge piping of the ore spray yetems, , HPCI, an RCIC a x ed: The ctioa of the RCIC d HPC pumps 11 aligned to the condeasa storage tank, d e pressure on chambe head tank shall normally be aligned to serve the discharge piping of the RHR aad CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC hea tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 peig 1-?4-65 48 psig Sr+Ycc+Q~,g~~<1ST5 3.g.)Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems shall be vented from the high point aad water flow determined. Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to serv 5 R 3.$.3.)3.Whenev RCIC system is lined up to take suction from the condensate stora e tank, the discharge iingof te RCIC s 1 e ven ed f m tge gh po t o the a'n at flo obs e cm s monthly as s.4.,When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.BFN Unit 1 3.5/4.5-17 NENOMENT NO, 2 06 PAGE~GP

UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP~w>*~

(g~PC/i gi~%ion 3~3 FEB 0 7 1991.5.E s Co a t ct o 4.5.E S ste C S ssu Co t'ect o S stem C S 4.5.E.1 (Cont'd)e.Flov Rate at Once/18 150 psig months CL Dv$4limftow d'or<ha-P~s 4'o~BPhl ISIS g,s.i f.Verify that each valve (manual, pover-operated, or automatic) in thc injection flov-path that is not locked, sealed, or othervise secured in position, is in its correct+position.Once/Month The HPCI pump hhall deliver at least 5000 gpm during each flow rate test.2.If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSSo RHRS(LPCI), and RCICS arc OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdovn shall be initiated and the reactor vcsscl pressure shall be reduced to 150 psig or less vithin 24*Except that an automatic valve capable of automatic return to its ECCS position vhen an ECCS signal is present may be in a position for another mode of operati l C<1.Thc RCICS shall be OPERABLE vhenever there is irradiated fuel in the reactor vessel and the reactor vessel~ppi'c4.i~~prcssure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 1.RCIC Subsystem testing shall be performed as follovs: Pc or LI 5'R3.5.3.5 a.mulated Auto-Once/18 matic Actuation months Test P.,~~Noh r sa>5.>>BFN Unit 2>~sM ie 3.s3.9 3.5/4.5-14 NENOMENr No.190 PAGE N OF

S CRJXiCOJAJOM NOV 24 1999 Prop@~M4e~sg g.s.g f be determined vithin 12 hour after reactor steam pressure reaches 150 psig from a COLD CONDITION or a ernat ve y RIOR~0 RRRR~Oy naROR an auxilia~steam supply.'A3 4.5.F.1 (Cont d)SR'3.5.5.3 b.Pump OPERABILI TY Specifi-cation 1.0.MM tor-Operat d Per Va e Spec+i-OPE LITT ation~1.O.MM s~s.s.3.5 d.HJf Ol+O~4 4 sR 8.5.3*2JRRO S Flov Rate t orma react r v sel bgcra~in re ure Once ct Zo 4o IOIO PSJ+2.If the RCICS is inoperable, the reactor may remain in Ac TloQ operation for a period not to exceed days if thc HP is OPERABLE durin such time,.ppp f,R J'~J~4JJ P 3.If Specifications 3.5.F.1 or 3.5.F.2 are not met,~5aQA~k and the reactor shall be depressurized to ss than 150 psig vithin-hours.N,~cg.4 SR 3.S:~'f e.c.5 sf'.S.X,3 SA 3.S.3.9 5'~>53.Z f.Qp,i Flov Rate at sig~/45 Thc RCIC pump shall deliver at least 600 gpm during each flow test.Once/18 months Verify that nce/each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-visc secured in position, is in its correc po ition.Except that an automatic alvc capabl of automatic r urn to its ormal pos tion vhen a signal is prese may be in position for another mode of operation. BFH Unit 2 3.5/4.5-15 AMENOMENT Na.I 76~Cji' ~' 5')Ci'ccrc'u 3.5:3 UG 02 1989 he core s ray system PC HPCI or RCIC are requ red to be OPERABLE, the scharge p ing from he pump d charge of hest syst s to the st bloc valve sh 1 be fille Sg 8.S.>.e, folloving surveillance requirements shall be adhered to assure that the discharge piping of the core spra systems HPCI and CIC are filled: The suc on of th RCIC and HPCI pumps sha be alig ed to the condensate torage tank, an e pressure suppress on chamber head tank shall normally be aligne to serve the discharge piping of the RHR and CS pumps.The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable. The pressure indicators on the discharge of the RHR and CS pumps shall indicate ot less than listed btlov.Pl-75-20 48 psig Pl-75-48 48 psig Pl-74-51 48 psig Pl-74-65 48 psi See X,q4<f'c~~io ~or C4 pg 4r 8P N ISTIC 2~g)gs a.)4.Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray system, the discharge piping of these systems shall be vented from the high point and vater flav determined. Folloving any period vhere the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the s stem ta servict Whenever the CI or RCIC system is lined up to take suction from the condensate storage tank, the discharge piping of the CI an RCIC e v ed fram t high po of the s tern and voter lov ob erve on a monthly basis.When t e an t e CSS are required to be OPERABLE, the pressure indicators vhich monitor the discharge lines shall be monitored daily and the pressure recorded.BFH Unit 2 3.5/4.5-17 AMEHbMENTHg. T 6 g PAGE~O~

UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP Cl ~<<~<'~ho B.S.3 FEB 0 7 l991 3.5.E essu e Coolant In ection S st PCIS 4.S.E i h ressure Coolant In ectio 4.5.E.1 (Cont'd)e.Flow Rate at Once/18 150 psig months 5 cd+iL5+jg'gg'on fia Ch+Qc5 Pnu BPn/]5y5 3 5.l Once/Month f.Verify that each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct*position.The HPCI pump shall deliver at least 5000 gpm during each flow rate test.2.If the HPCI system is inoperable, the"reactor may remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.2.No additional surveillances are required.3.If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.*Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation. Lco X5.3%Plica'ikj The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.OPERABILITY shall 1.RCIC Subsystem testing shall be performed a follows: A~or SP,g,g3.5 a.Simulated Auto-Once/18 matic Actuation months Test nap A h4w~sl s.S.~.s BFN Unit 3<>Rsed.~Q~Spy g 3 3.5/4.5-14 AMBDMEHTNO, yg p S HS FOR OPE s+~<;~+~-~NUV a+ious be determined vithin 12 hour t 4f'~~after reactor steam pressure reaches 150 psig from a COLD SC3.g,3.9 COHDITIOH r a t at e y PRI TO S RTUP usi an auxil st am sup ly.5'gg g p p b Pump OPERABILZ1Y .M tor-Operate Va e OPE ILITY er Specifi ation.O.MM r Sp cifi-cat n 1.0.Sg 3,g,pp d.rpo5rd go&sP.w,s;z.s SR Z.S.P.q l.6<CrS,a.>Sg3 Flov Rate at Once orma rea to v sel oper ting pr sur zo+o fo io t'so c)Flov Rate at Once/18 sig months C rgb The RCIC pump shall deliver at least 600 gpm during each flov test.2.If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed days if the Jq HPCIS is OPERABLE durin such time.ripe/'~md'atcl L.z 1 3.If Specifications 3.5.F.1 or 3.5.F.2 are not met, ee-D~~d-and the reactor shall be depressurized to less than 150 psig vithin 2A hours.3L>R'ua,'l H Slt'R 5.xz f.Br'n~3 ih f2hc g Verify that'ce/each valve (manual, pover-operated, or automatic) in the in)ection flovpath that is not locked, sealed, or other-vise secured in position, is in its correc sition.s*cept hat a aut atic v ve c able f aut matic re urn t its n rmal po tion en a igna is pre ent ma be in posi ion fo anoth r mo of operation. J~BFN Unit 3 3.5/4.5-15 AMENDMENT HO.1 4 4

~+4 Pmfion 3.5.3 NY i 9 894 WAmeovos-he core spray systems CI o C C are equired to b OPERAB E, t disc arge L.Al pipi from t e p disc arge of th se syst s to the la t block lve s ll be ille~RE.The following surveillance requirements shall be adhered to assure that the dis harge pipin of'he core spray systems, LPCI, HPCI, and RCIC are x e e sucti n of th R~C and HPC p s shal be ali ed to t e con sate s ra e t k the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.The condensate head teak may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.. The pressure indicators on the discharge of'he RHR and CS pumps shall indicate not lees than listed below.20 Every month and prior to the testing of the RHRS (LPCI and Containment Spray)and core spray systems, the discharge , piping of these systems shall be vented from the high point and water flow determined. Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system service.Pl-75-20 Pl-75-48 Pl-74-51 Pl-74-65 48 psig 48 psig 48 peig 48 psig S~RB.5.3 3 Whenever the PCI RCIC system is lined up to take suction from the condensate storage tank, the discharge pi iag of the CI an RCIC gee 3'uSb4ecaW< fpr Q)g,~~~For Sg'H 4 5g p e yea e rom g e bligh poin of Qe s nK wate low o serve on a monthly bas s.4.When e RHRS and the CSS ar required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.BFN Unit 3 3.5/4.5-17 NENOMENT tt0.I 7 B PAGE~OF~ JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 The Frequency of"Once/month" has been changed to"31 days." The Frequencies of"Once/3 months" and"Per Specification 1.0NH" have been changed to"92 days." Since the proposed frequencies are equivalent, this change is considered administrative. A3 Notes allowing actual vessel injection to be excluded from this test (simulated automatic actuation test)have been added to proposed SR 3.5.3.5.Since the current requirements state the test is"simulated" (i.e., valve actuation and vessel injection are inherently excluded), this allowance is considered administrative in nature.A4 Surveillance Requirements for HOV operability that are required by the Inservice Testing Program have been removed from individual Specifications. This change is considered administrative in nature since these requirements remain in the IST Program which is defined by proposed Specification 5.5.6.A5 The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system.Since the reactor steam dome pressure must be&20 psig to perform SR 3.5.3.3 and&50 psig to perform SR 3.5.3.4, sufficient time is allowed after adequate pressure is achieved to perform these tests.This is clarified by a Note in both SRs that state the Surveillances are not BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM required to be performed until 12 hours after the specified reactor steam dome pressure is reached.CTS 3.5.F.1 already contains the context of the Note for the low pressure flow rate test.This is also consistent with interpretation of the current technical specification requirement for the high pressure flow rate test which is currently not modified by a Note.A6 The clarifying information contained in the"*" footnote has been moved to the proposed Bases for SR 3.5.3.2.The intent of the surveillance is to assure that the proper flow paths will exist for RCIC System operation. The Bases clarifies that a valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.Moving this clarifying statement to the Bases is considered administrative in nature.A7~A8 A finite Completion Time has been provided to verify HPCI OPERABILITY. the new time is immediately and is considered administrative since this is an acceptable interpretation of the time to perform the current requirement. CTS 3.5.F.3 requires the reactor to be depressurized to less than 150 psig when CTS 3.5.F.1 and 2 cannot be met, while CTS 3.5.F.1 requires RCIC to be OPERABLE when reactor vessel pressure is above 150 psig.Proposed Required Action B.2 requires the vessel to be depressurized to x 150 psig.Since the intent of CTS is the same even though the CTS shutdown statement does not state"equal to," the addition of this requirement is considered administrative. TECHNICAL CHANGES-MORE RESTRICTIVE An additional requirement is being added that requires the plant to be in MODE 3 within 12 hours.This change is more restrictive because it stipulates that the reactor shutdown be completed much earlier than would be required by the existing specification (CTS 3.5.F.3).CTS require a shutdown to<150 psig within 24 hours but do not stipulate how quickly NODE 3 must be reached.Reference Comment L3 which addresses the less restrictive change of being~150 psig in 36 hours rather than 24 hours.BFN-UNITS 1, 2, 5 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" LAl The details relating to system design and purpose have been relocated to the Bases.The design features and system operation are also described in the FSAR.Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the FSAR will be controlled by the provisions of 10 CFR 50.59.System operability determination, as described in the Bases and SR 3.5.3.1, will ensure maintenance of filled discharge piping.LA2 The details relating to methods of performing surveillance test requirements have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs.LA3 CTS 3.5.F.1 specifically states that RCIC Operability can be determined prior to startup by using an auxiliary steam supply in lieu of using reactor steam after reactor steam dome pressure reaches 150 psig.Details of the methods of performing this surveillance test requirement have been relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled programs."Specific" Ll The phrase"actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying that the RCIC System actuates on an automatic initiation signal.This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adequately demonstrated in either case since the RCIC System itself can not discriminate between"actual" or"simulated." L2 BFN-UNITS 1, 2, 5 3 This change proposes to extend the current allowed outage time for the RCIC System from 7 days to 14 days.The 14 days are allowed only if the HPCI System is verified Operable immediately. Loss.of the RCIC System will not affect the overall plant capability to provide makeup inventory at high reactor pressure since the HPCI System is the only high pressure system assumed to function during a LOCA.However, the RCIC System is PAQE J pF 3~~Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM the preferred source of makeup for transients and certain abnormal events with no LOCA (RCIC as opposed to HPCI is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of the RPV water level).The 14 day completion time is also based on a reliability study that evaluated the impact on ECCS availability (Memorandum from R.L.Baer (NRC)to V.Stello, Jr.(NRC),"Recommended Interim Revisions to LCOs for ECCS Components," Oecember 1, 1975).Because of similar functions of HPCI and RCIC, and because HPCI is capable of performing the RCIC function, the allowed outage times determined for HPCI can be applied to RCIC.This change is consistent with NUREG-1433. L3 The time to reduce reactor steam dome pressure to a 150 psig has been extended from 24 hours to 36 hours.This provides the necessary time to shut down and cool down the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE.This extra time reduces the potential for a unit upset that could challenge safety systems.In addition, a new (more restrictive) requirement to be in MODE 3 (Hot Shutdown)within 12 hours has been added (See Comment Ml above).These times are consistent with the BMR Standard Technical Specifications, NUREG 1433.L4 Existing Surveillance Requirement 4.5.E.l.d requires verification that RCIC is capable of delivering at least 600 gpm at normal reactor vessel operating pressure.The proposed surveillance, SR 3.5.3.3, requires verification of a minimum 600 gpm RCIC flow rate with reactor pressure~920 psig and<1010 psig.The RCIC performance test at high pressure is the second part of a two part test that verifies RCIC pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the RCIC pump is expected to perform.Performance of the RCIC test at both ends of the expected operating pressure range confirms that the RCIC pump and turbine are functioning in accordance with design specifications. The ability of the RCIC pump to perform at normal reactor vessel operating pressure has already been demonstrated. A small decrease in the pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. L5 Existing Surveillance Requirement 4.5.F.l.e requires verification that RCIC is capable of delivering at least 600 gpm"at 150 psig reactor steam pressure." The proposed surveillance, SR 3.5.3.4', requires verification of a minimum 600 gpm RCIC flow rate with reactor pressure BFN-UNITS 1, 2, 5 3 Revision 0 ' JUSTIFICATION FOR CHANGES BFN ISTS 3.5.3-RCIC SYSTEM at 165 psig.This change is less restrictive because it could allow reactor operation at pressures up to 165 psig prior to performing the surveillance. Performance of RCIC pump testing draws steam from the reactor and could affect reactor pressure significantly. Therefore, RCIC pump testing must be performed when the Electro-Hydraulic Control (EHC)System for the main turbine is available and capable of.regulating reactor pressure.Operating experience has demonstrated that reactor pressures as high as 165 psig may be required before the EHC system is capable of maintaining stable pressure during the performance of the RCIC test.The RCIC performance test at low pressure is the first part of a two part test that verifies RCIC pump performance at the upper and lower end of the range of steam supply and pump discharge pressures in which the RCIC pump is expected to perform.Performance of the RCIC test at both ends of the expected operating pressure range confirms that the RCIC pump and turbine are functioning In accordance with design specifications. The ability of the RCIC pump to perform at the lowest required pressure of 150 psig has already been demonstrated. A small increase in the pressure at which the performance to design specifications is verified will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications. L6 CTS 3.5.F.1 requires operability to be determined within 12 hours after reactor steam dome pressure reaches 150 psig from a COLD CONDITION. The allowance for reactor steam dome pressure and flow to be adequate is based on the need to reach conditions appropriate for testing.The existing allowance to reach a given pressure only partially addresses the issue.This pressure can be attained, and with little or no steam flow, conditions would not be adequate to perform the test-potentially resulting in an undesired reactor depressurization. The proposed change recognizes the necessary conditions of steam flow and minimum pressure as well as a maximum pressure limitation and provides consistency of presentation of these conditions. The point in time during startup that testing would begin remains unchanged. The change simply changes when the 12 hour clock for performing the test must begin and permits testing to be completed in a reasonable period of time.BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES BFN ISTS 3.5-ECCS AND RCIC SYSTEM BASES The Bases of the current Technical Specifications for this section (3.5.A, B, E, F, G, H, and 4.5)have been completely replaced by revised Bases that reflect the format and applicable content of proposed BFN-UNIT 1, 2,'and 3 ISTS Section 3.5, consistent with NUREG-1433. The revised Bases are as shown in the proposed BFN-UNIT 1, 2, and 3 Bases.BFN-UNITS 1,'2, 5 3 pAGF l Revision 0 UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP J S pdCiliCcyhon 7,C, (,/FEB 2 ames Ezo gy,/,]"/Pl<'~nb'ily erforming"open vessel" physics tests at pover levels not to exceed 5 MW(t).A b.Primary containment integrity is con rmed if e maxim allov le in egrated eakage rate, Lay does no't exceed the equi lent of pere t of the pr ry co ainmen volume r 24 h rs at the 49.6 psi design asis accident essure P'.Ce If 2 makeup to the rimary ontai ent ave aged ver 2 hours (correc ed f pr ssure, tempera ure, d ven ing op ations exc ds 542 CFH, it must b red ce to<42 SC vithin~ours<gioNpg or the reactor shall be emplaced in Hot Shu de RGTl&t 6)vithin the next 3k hours~2 2.a.Primary containment 4'/~~Pl'4 4aca~y shall be maintained at all times vhen the reactor is critical or vhen the reactor vater temperature is above 212 F and fuel is in the reactor vessel cep v e Primary co tainment n trogen consumpti n shall be monitor to etermin the averag dail nitrog cons tion or the ast 24 h urs.cessi leakage is ndicat d by a co umpti n rate f>I of e pr ry con ainm t free lume er 24 ours corre ed f dryv ll temper ture press re, venti op atio)at 49.6 psig Corr cted t no 1 d ell perati pressur of 1.psig, this value s 542 CFH.this value is exc eded, e action spec fied in 3.7.A.2.C shall be taken.5'R3-4~~~.<in accordance vith the Primary Containment Leakage Rate Testing Program.~4<<id'hu>~~n34ho, s BFK Unit 1 3.7/4.7-3 NENOMEg gg, pp8 FE822~i PAGE~OF 3.7/4.7W DF<c4'c'phon 3',(., l, 1 FEB 8 2 199j Pr,(.cour~ BFS Unit 1 3.7/4.7-5 hMENOMENT NO.2 2 8

Yush4caÃon gz PQ~y 0<8gaJ]Spy~~~<R 3.R.4~I'e I g.Perform required local leak rate tests, nc u ng t e r mary containment air lock leakage rate testing n accor ance v t t e Primary Containment Leakage Rate Testing Program.Eote: An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFH Unit 1 3.7/4.7-6 AMENDMENT NL?P.8 (1)Ef at ny time it is termined hat t criteri of 4.7.A.2.g's exceeded, repairs shall be initiated immediatel in Hob'in/s.0c~rs I~>><'/>n gg A~~pc.T(oN A (2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within~i hours.following detection of excessive local leakage, the reactor shall be until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by etest.The mann s earn one isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 3.7/4.7-8 When e prima conta ent's i ert the con inment s ll continuo ly monit red for gross eakage b viev of e inerti reg rements.This monit ing sys may be aken o t of se ice for nt e but s 1 be ret rned t service as soon pra icable.The in rior su aces the d Il and t ab e the evel one foot belo the n anal vat line out de surfac of th torus elov th vater ine 1 be suall ins cted ope sting cle or de riorat on signs st ctur e vith par cul attention to pip conn tions and suppo s and or signs of dist ess or displacement. BPS Uait 1 3.7/4.7-9 3.7.k.4 (Cont'4)4.7.4.4 (Coat'4)c.so Cryvell-suppression chamber vacuum breakers may'oe determined to be yerable for opening.Scc W Chewy>W B Ai~F5 3AI.i d<<<<f Spec<<<<cac<<oas 3~7.k<<<~aq 3.7.4.4.b, or 3.7.4.4.c. cannot be met, the aait shaLL be placed in a COLD SHUTDOWNÃCQ5D~ON in an orderly maaaer vithin 24 hours."-ach vacuum breaka valve shall be inspect~for yroyer oyeratioa oi="e valve aaC LM in accordance vith Spec'ficat'on 1.Q..C.Q~gi~le li 2-a tesc of the dr:ovel to suppression chambe st~care shall be conducted daring each o Lcc c Le Leaic rate Ls la~og.0.09 Lb/sec M yr~~containment acmosyhere vi" 1 ysi Cifie"mat'al 5.a.Containment atmosphere shall be reduced to Less than 4X oxygen vich aitrogea gas Curing reactor yover oyeration vith reactor cooLant yressare above 100 ysig, except as specified in 3~7.l<<5<<b.a.The yrimary.containment oxygea concentration shall be measured and recorde4 daily.The oxygea measurement shall be ad)usted to accoaat for the aacertainty of the metho4 used by adding a predetermined error faact'oa.b.Vichia the 24 hoar period subsequent to ylacing the reactor in th>>RN NDl foLloving a shat-Cova~the coatainmeat atmosphere oxygen coacentracioa shall be reduced to Less than 4%by volume and maintained Ln this condition. Deinerting may commence 24 hours prior to a shatCova~b.The methods used to measure the primary containment oxygen coaceatration shall be calibrated once every refueling cycle.c.If plaat control air ts being used to supply the yaeamatic coatrol system-inside primary coatainmcat, the reactor shall aot ba started, or Lf at yover, the reactor shall be brought to a COLD SHUTDOWN CQHDIT 05 vithia 24 hours.c.The coatroL air suyyly valve for the pneumatic coatroL systes Laside the pzmaz7 coatainmeat shall be verif'ea i closed prior to reactor star==and monthly thereafter. d.If Specificatioa 3.7.A.5.a aad 3.7.4.5.b cannot be met, aa orderly shutdova shall be initiated aa4 the reactor shall be in a COLD SHUTDOWN COHDITIOI vithia 24 hours.BFH Unit 1 3.7/4.7-11 AMEi fOMENT NQ, y g g PAGE I OF i UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP Cthe*Op, Qe J~J 2eae&0 3.k.i,J A p)i cg b i li 4 b.C~Alarm A Wlnpl 8 Primary containment maintained at all times when the reactor is critical or when the reactor water temperature is above 212 F and fuel is in the reactor vessel cept w er orming"open vessel" physics tests at power levels not to exceed 5 M(t).2 Primary conta nmen ntegri is confirme if t e max allo ble in egrate leakag rat , La, does n t excee th egu alent f 2 pe ent f the imary ontai t volum per 24 ours the 49.6 p ig desi basis ccident pressure, P.If 52 makeup to the primary con ainmen averag d over 24 urs (c recte for pres ure, t eratu e, and venti opera ons)ceeds 542.S, it uced to SC ithinA ours/or the reactor shall be placed in Hot Shutdown within the next hours Sf'rimary contai ent nitrogen consum tion s 1 be monit ed to d termine he aver e daily itrog cons ption f r the 1 st 24 ours.cessive eakage is ndicate by a H co umptio rate of>2X of t e prima contai ent free olume pe 24 hou (correct for d ell tempera re, pre sure, venting operati ns)at 49.6 p ig.Co rected to norma drywel opera ing pres re of.1 psi , this val is 54 SCFH.If this val e is ceded, e act on specified in 3.7.h.2.C shall be taken.~7.C.(.l e I er rm leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program d~hl Shu~a'~36 hOl4CS BFH Unit 2 3.7/4.7-3 hNENMENr NL 2 c 8

Sp<c jfirqQn Zi io.I>/FEB 8 259s BFR Unit 2 3.7/4.7W NENDMENT NC.2 4 3 PAGE~o..'I 5f ecsP:c'crgon 9.6, (.I FES 2 2 1996 4~~e t BFE Unit 2 3.7/4.7-5 hMENOMENT NO.2 4 3

FEB 22$9S~~~~A/3 3.4.lil g.Perform required local leak rate tests nc u ng e primary containment air lock leakage rate testi in accordance w e Primary Containment Leakage Rate Testing Program.See ruse<'~Son 4r c+~$Ar/go zsTs 3.o, I.~Hote: kn inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFH Unit 2 3.7/4.7-6 NENDMENT No.2 4 3 5

h: (1)If at any'me it is determi that e crite on of 4.7.2.g Is exceeded, repairs shall be initiated immediately. r~HojX-: g i~l2 N~r~+>>DC q;3g 4o~rg (2)CT(ofJ p,~fA p4 l AC7)o~B If conformance to the criterion of 4.7.A.2.g is not demonstrated within hour&f oil owing detection of excessive local leakage, the reactor shall b until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling'utage.If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. ~kQ MKgPJ ji (g)li~g~~>+>isrs g.S./g BFN Unit 2 3.7/4.7-8 PAGE j.o HmDar en the pr ry c tainment incr the ontainmena, shall be cont uously mo tored fo'r gr s leakage by reviev o the incr stem up r uircmen~This ear may bc taken ut of~rvice for maint a but be returned o serv ce as soon as pr ticable.e interior fac s of e dzpwell to ab e the leve one f t bel the no eater line outside surfac of the t belcnr eater 1 shall be sually pected ch opera iIlg le for de rioration and signs struc al e vlth parti ar atten ion to ping acti and s rts for signs o'f d tres or displac t BPI Unit 2 3.7/4.7-9

S'7 HOV 22 1888 5 acifica4iow E Co (\S 3.7.A.4 (Cont'd)4.7.A.4 (Cont d See v~dÃicJ~CQCs+AQ SAI isTS g.to I.7 c.Tvo dryvel'-suppression chamber vacuum breakers may be determined to be inopc abLe for opening.d.If Specifications 3.7.A.4.a.b, or.c cannot be met, thy uait shall be placed in a Cold Shutdovn condition in an orderly manner vithin 24 hours.SEE KcCSr tFiCArcrg~g CAAHQ~>R QFN<srs g.c,g~0 c.Each vacuum brcake" valve shall bc inspected'or proper operation of="e valve and limit switches in accordance with Specification L.O.HC.gp q.Q.).~d.A leak test of the d Jvell to suppression chamber tructure shall be conducted during each P l Accc table e rare~pg, 0.09 1 scc o pr mazy coatainmcat atmosphere with si differential a.Coatainmeat atmosphere shall bc reduced to less thaa 4X oxygen vith nitrogea gas during reactor pover operation Mich reactor.coolant pressure above 100/psig, except as specified ia 3.7.A.S.b. a.The primary coataiameat oxygen concentration shal'e measured and recorded daily.The oxygen measurement shall bc ad)usted to account"for the uncertainty of the method used by adding a predetermined error function.b.Mithin the 24-hour period subsequent to placing the reactor in the RUH mode folloving a shut-down, the containmeat atmosphere oxygen coaceatration shall bc reduced to less than 4X, by volume and maintained in this condition. Deinerting may commeace 24 hours prior to a shutdown.b.The methods used to measure the primary containment oxygen concentration shall be calibrated once every refueling cycle.c.If plant control air is being used to supply the pneumatic control system inside primary coatainment, the reactor shall aot be started, or if at pover, the reactor shall bc.brought to a Cold Shutdovn condition vithin 24 hours.c.The control air supply valve for the pneumatic control system inside the primary containment shall be verified closed prior to reactor startup and monthly thereafter. d.If Specification 3.7.A.S.a and 3.7.A.S.b cannot bc met, an orderly shutdovn shall be initiated and the reactor shall bc in a Cold Shutdown condition vithin 24 hours.BFH Unit 2 3.7/4.7-11 ph3c'gMg!T t'~C PAGE~OP ' UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP 5 Pc 4 i Ci cp on, g]2.a.Primary containment maintained at all times whea thc reactor is critical or when the reactor vatcr temperature is above 212 F and fuel is in the reactor vessel ccpt w i e pcrformiag"open vessel" physics tests at pover levels not to exceed MW t b.Primary ontaiamcnt i tcgrity s confi ed if th maxim allovab e in grated eakagc r te, La, does not exceed t e equi alent of 2 percen of the p imary co taiament volume er 24 h urs at t e 49.6 ps g design basis ccident P~c If H2 makeup to e primary c ntaiam t aver d over 24 hours (orrected or pr sure, t peratur and vent ng opera ions)cx eds 542 S , it t bc reduce to<5 SC vithin hours g or the reactor shall be c, placed in Hot Shutdovn'8 Swithin the next hours./2 Primary c taiament nitrogen coasumpt n shall monitor d to dete ine thc averag daily,ni rogen cons ption fo the las 24 urs.Ex essivc 1 ge is ndicate y a N2 c umptio rate of 2X of e prima conta ent frcc volume r 24 hou (corrc cd for ell'cmpe ature, p ssure, vcn ag opera ioas)a 49 psig.orrecte to n 1 d ell oper iag pressure f 1.1 ps g, this value is 542 SCFH If this value i excccd , thc action specifi in 3.7.k.2.c sha be taken.S 34~er leakage rate testing in accordance vith the Primary Contaiamcnt Leakage Rate Testing Program.~told Shukdoupn>n 34 goer g BFS Unit 3 3.7/4.7-3 lRNMNT NO.2 03 PAGE OF Sk<'4 0~Z<FEB 2 8 SSS BFR Unit 3 3.7/4.7W NIENDhfQT NQ, P 03 PA3E~OF~ eted BPH Unit 3 3.7/4.7-5 NBIDMEÃF RLP 0 3 PAGE~OF

g.Perform required local leak r eats nc u ng t e primary containment a r lo e rate testi in accordance v t t e Primary Containment Leakage Rate Testing Program.5ee 3~g40;rabin&<Ch~qcs&r SAN Xsrs Zt..l.<Rote: An inoperable air lock door does not invalidate the previou successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)@hen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).BFR Unit 3 3.7/4.7-6 NBIOMENT HO.2 03 PAaE h.(1)If at a time it is d ermined that the criterion o 4.7.A.2.g is exceeded, repairs shall be 3.nitiated immediately. pl~3 j~/2&Owed~Ho05 Q/~2C,A-<<(2)f conformance to the criterion of jhow'IorJ A 4.7.A.2.g is not demonstrated within~hour>following detection of excessive local eakage, the reactor shall be until+vta4 8 repairs are effected and the local leakage meets the acceptance criterion as demonstrated by etest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. 5j?t j~g jiCiGaf>ow 4~C44 p+AI BFN l>7-s g.z./.3 BFN Unit 3 3.7/4.7-8 PAGE

C t the pr ry con ainment i inerte thc ntainment hall contin usly mo ored for gros leakage eviev of e inert tea aike r remcnts.This monit rial sys may bc taken t of sc ce for maint e but shall be returned t service as soon as yrac icable.The interior surfaces of the dryvcll and torus above the level onc foot belov the normal vater line and outside surfaces of the torus belov the water linc shall be visually inspected each operating cycle for deterioration and any signs of structural d4RLgc vith particular attention to pipiag connections and.suyports and for signa of dis'tress or displacement. BPH Unit 3 3.7/4.7-9 0 3.7.A.4 (Cont'd)4.7.A.4 (Cont'd)c.Tvo dryvcll-suppression chamber vacuum breakers may be determined to be inoperable for opening.d.Sec'$45f&~og For chpggs+r 8cN fsrs 34(,7 If Specifications 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c, cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner vithin 24 hours.C~d.Once each operating cycle, each vacuum breaker valve shall be inspected for proper operation of the valve and limit svitches in accordance with Specification 1.0.MM.A leak test of thc dryvell to suppression chamber structure shall be conducted ur ng a o Acce table leak rate is.09 lb/sec o pr ary containment atmosphere vit 1 si differential. 5.0 5.0 a.Containment atmosphere shall be reduced to less than 4X, oxygen vith nitrogen gas during reactor paver operation vith reactor coolant pressure above 100/psig, except as specified in 3.7.A.5.b. a.The primary containmcnt oxygen concentration shall be measured and recorded daily.The oxygen measuremcnt shall be ad)usted to account for the uncertainty of the method used by adding a predetermined error function b..Within the 24-hour period subsequent to placing the reactor in the RUR mode folloving a shut-down, the containment atmosphere oxygen concentration shall bc reduced to less than 4X by volume and maintained in this condition. Deinerting may commence 24 hours prior to a shutdown.b.The methods used to measure the primary containmcnt oxygen concentration shall be calibrated once every rcfucling. cycle.c.If plant control air is being used to supply the pneumatic control system inside primary containmcnt, the reactor shall not be started, or if at pover, the reactor shall bc brought, to a Cold Shutdown condition vithin 24 hours.If the specifications of 3.7.A.5.a through 3.7.A.S.b cannot be mct, an orderly shutdown shall bc initiated and thc reactor shall bc in a Cold Shutdown condition ithin 24 hours.c.Thc control air supply valve for the pneumatic control system inside the primary containmcnt shall be verified closed prior to reactor sr.artup and monthly thereafter. Ce 3'u5hCs'cation fir<~g<f~BPu~5'r>Z.G.3.2.p~cE~oF 0 BFH Unit 3 3.7/4.7-11

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIMARY CONTAINMENT ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. A2 Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.The definition of PRIMARY CONTAINMENT INTEGRITY has been deleted from the proposed Technical Specifications. In its place the requirement for primary containment is that it"shall be OPERABLE." This was done because of the confusion associated with these definitions compared to its use in the respective LCO.The change is editorial in that all the requirements are specifically addressed in the proposed LCO for the primary containment along with the remainder of the LCOs in the Containment Systems Primary Containment subsection (e.g., air locks, isolation valves, suppression pool).Therefore, the change is purely a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433.A3 CTS 4.7.A.2.k requirements for visual inspection of the drywell and torus surfaces are also contained in 10 CFR 50, Appendix J.These regulations require licensee compliance and cannot be revised by the licensee.These details of the regulations within CTS are repetitious and unnecessary. Therefore, the details also found in Appendix J have been deleted.This is considered a presentation preference and as such is considered an administrative change.BFN-UNITS 1, 2, 8L 3 Revision 0 PAGE~OF

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIHARY CONTAINHENT A4 A5 CTS 3.7.A.2.b provides acceptance. criteria for integrated leak rate testing, which is redundant to those contained in Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3)requirements. The definition of L.is provided in proposed BFN ISTS 1.1 and need not be repeated here.As such, this deletion is considered administrative. The acceptance criteria for the leak test of the drywell to suppression chamber structure has been changed from 0.09 lb/sec of primary containment atmosphere at 1 psid to 0.25 inches of water for 10 minutes.Since these values are equivalent this is considered an administrative change.A6 CTS 4.7.A.2.h(1) requires repairs to be initiated immediately when it is determined the criterion of 4.7.A.2.g is exceeded.CTS 4.7.A.2.g requires LLRTs to be performed in accordance with the Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3).CTS 4.7.A.2.h(2) then allows 48 hours to demonstrate 4.7.A.2.g can be met following detection of excessive local leakage.Since repairs are typically initiated immediately and proposed BFN ISTS ACTION A will only allow 1 hour"to restore primary containment to OPERABLE status prior to requir'ing the initiation of a shutdown (reference Justification H2 below), CTS 4.7.A.2.h(1) has been deleted.TECHNICAL CHANGES-NORE RESTRICTIVE Hl CTS 3.7.A.2.a requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.1 applicability is HODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in HODE 2, not critical and<212'F.H2 Proposed Action A is more restrictive than CTS 3.7.A.2.c since the time allowed to reduce excessive nitrogen leakage prior to initiating a shutdown has been reduced from 8 hours to 1 hour.The time allotted to place the unit in Hot Shutdown (HODE 3)has been reduced from 16 hours to 12 hours.Proposed Action B requires the unit to be placed in Cold Shutdown (HODE 4), whereas, CTS 3.7.A.2.c only requires the unit to be placed in Hot Shutdown.In addition, CTS 4.7.A.2.h.(2) allows 48 hours to demonstrate conformance to Appendix J following detection of excessive local leakage BFN-UNITS 1, 2, 5 3 Revision 0 PAGE R op 3

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.1-PRIMARY CONTAINMENT and then requires a plant shutdown if conformance can not be demonstrated. CTS does not specify a completion time for shutdown and does not specify whether shutdown is to the Hot or Cold Shutdown Condition. The Proposed Actions A and B are more restrictive since they only allow 1 hour to restore primary containment and then require the unit be in MODE 3 in 12 and MODE 4 in 36 hours.TECHNICAL CHANGES-LESS RESTRICTIVE"Generic" programs.LCl The conti nuous leak rate monitor does not necessarily relate directly to primary containment operability. In general, the BWR Standard Technical Specifications, NUREG 1433, do not specify indication-only or alarm-only equipment to be OPERABLE to support operability of a system or component. Control of the availability of, and necessary compensatory activities if not available for, indications, monitoring instruments, and alarms are addressed by plant operational procedures and policies.Therefore, the continuous leak rate monitor, and associated alarm surveillances and actions will be relocated to a licensee controlled document.Any changes will require a 10 CFR 50.59 evaluation. LA1 The details relating to routine monitoring of plant status and operations parameters that reflect primary containment operability and the methods of performing this monitoring have been relocated to the Bases and procedures. Acceptance criteria for primary containment N, leakage (i.e., makeup consumption) have been relocated to procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in proposed BFN ISTS Section 5.0 and changes to the procedures will be controlled by the licensee controlled BFN-UNITS 1, 2,&3'AGE QP~Revision 0

UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP 0 SAci 0i c~FEB 2 2 1996 At 2.a.Prima conta nment inte ity s ll be mai aincd t all imes vh the cactor is cr ical o vhen e re tor v ter cmpcr urc i abov 21'P and f el is n the ea or vcss cxc t wh e perf rmi"op vcs el" physics sts a po er levels not to exceed 5 t%(t).b.Primary containment integrity is confirmed if thc maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of.thc primary containment volume per 24 hours at thc 49.6 psig design basis accident pressure, Pa.c.If H2 makeup to the primary containment avcragcd over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must bc reduced to c 542 SCFH vithin 8 hours or thc reactor shall be placed in Hot Shutdovn vithin the next 16 hours.2.te ated Leak Rate estf Primary containment nitrogen consumption shall bc monitored to determine the average daily'itrogen consumption for the last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCPH.If this value is exceeded, the action specified in, 3.7.L.2.C shall be taken.Perform leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.gee guc+4'mb'+n Ar Chanyeg 4i BF~iSVS 3.C..i.l l-co p,5,1,2.8 lic4b.lip PI<RsR ACT'Iow5 A+g A]t'<opsy AJoH gg~/b;17oAS W3'Af$4 SR 3,g, t, f BFH Unit 1 3.7/4.7-3~AGE OF Sce XusHC'cab'on Qr@~ay'&<B~W tST5~~~~SR3.a.(.g. I g.Perfo required local leak rate tests including t e primary containment air lock leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program.Rote: An inoperable air lock 5g 3 g i g door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)when tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)when the air lock is pressurized to (g 2.5 psig for at least 15 minutes).<<e~~Wi'ca~n Qi C4~s 8FN)ST$5,g,I~BFH Unit 1 3'/4.7-6 AMENDMENT NQ.2 2 8 PAGE~QF g. Zf at any ime i" is det ined t at the iterio of 4.7.A.2.g exceeded, repairs shall be initiated immediately. ~~~3 I I24~/Hop/c/g g ggg within-48-ho rs f ollowing 2 4 HZ detection of excessive local leakage, the eactor shall be until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.(2)Ef conformance to the criterion of 4;7.A.2.g is not demonstrated The main steamline isolation valves shall be tested at a prgssure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 P.7/4.7-8 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP

SPCCi4iCCrf)On "7 6-I-2 FE8 P, 2%6 2.a.Primary ontai ent integri y sha be mainta ned a all times vhen e re tor is critical I or v en th react vat r t cratu is a ve 2 2'F fuel s in e re ctor v esel cept ile pcrfo ng"op ves el" physic tests at po er level not to exceed b.Primary containment integrity is confirmed if the maximum allovable integrated leakage rate, La, docs not exceed the equivalent of 2 percent of thc primary containment volume per 24 hours at thc 49.6 peig design basis accident pressure, Pa.c.If S2 makeup to thc primary containment averaged over 24 hours (corrected for'rcssure, temperature, and venting operations) excccds 542 SCFH, it must be reduced to<542 SCFH vithin 8 hours or the reactor shall bc placed in Hot Shutdovn vithin the next 16 hours.2.Pr'imary containment nitrogen consumption shall be monitored to dctcrmine the average daily nitrogen consumption for thc last 24 hours.Excessive leakage is indicated by a E2 consumption rate of>2X of the primary containment free volume per 24 hours (corrcctcd for dryvell temperature, pressure, and venting operations) at 49.6 peig.Corrected to normal dryvell operating pressure of 1.1 peig, this value is 542 SCFH.If this value is exceeded, the action spccificd in 3.7.k.2.C shall be taken.Perform lcakagc rate testing in accordance vith thc Primary Containment Leakage Rate Testing Program.~<~~~iWWn 4<C~C5 A~8FhJ ISTIC ACO g, (o.t.2 Ppp))~g;)', P f'no@i~<RCT<o~a/I>b Pno sea 4u4 I tg p Acr<o+5]go g<J 5k'.6.lit 2 BFH Unit 2 3.7/4.7-3 hMENMNr N.2 c 8 PAGE R OP~ Ai I:ES 8 2896 5<c 5~~$icqgan Qr C/ggg5 Qr SPN l5i5 Z.6, l.l~~~~58 3.b.I~>~~g.Perform required local leak rate tests, includi the primary containment air lock leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.Hote: kn inoperable air lock door does not invalidate the previous successful performance , of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).Se'e Su5 tea'can'en Qr'kCi~gC'5 Ai 8<N l57<g.g.i~BFH Unit 2 3.7/4.7-6 NENDMENT N.243 pAGE X o~~ ~r w>>n a.>>>,/h.Ql)Zf at y time it i dete ined that the cri erion of 4.7.A.2.g i exceeded, epairs shall be initiated immediately. In~p~3~2 He>>rj p go~g~4 haul Rpgu>l cA Acfi'p<<C.2.P~pa~RCy<<>>Q Ac.f>~C.I P,c, ripe b (2)f'onformance to the criterion of 4.7.A.2.g is not d strated within ours following detection of excessive local leakage, the reactor shall b shet-down until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Ef the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. SeC Xug$,'f;~p~gpg hr 5/WIST~36 1..3 BFN Unit 2 3.7/4.7-8 PAGE~0~ 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 2.a.Prima ontainment int ity shal e ntained all times hen th eactor is itical or v the rea r vater t'rature above 2 F fuel in the actor vessel cept v e perf ing"o vessel" p sics tes at pover evels no to exceed%l(t)b.Primary containment integrity is confirmed if the maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.c.If H2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to<542 SCFH vithin 8 hours or the reactor shall be placed in Hot Shutdovn vithin the next 16 hours.2.I e rated Leak Rate Testi Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for the last 24 hours.Excessive leakage is indicated by a E2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal dryvell operating pressure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified in 3.7.k.2.c shall be taken.Perform leakage rate testing in a'ccordance vith the Prima Containment Leakage Rate Testing Program.5ee~~+;4i ca fjon4r Chang ts'~8@v 1ST'.C;I.I ~c.g.b, l.2 I;i~b:l&t'ro scd 4'TYPES 8 FB~>+<<g Sole Jyg y+'o<S%3 fno acA, Sk 3,k.l.l.BFS Unit 3 3.7/4.7-3 NBfDMNT go.P g 3 f'ACiE~QF~

5'cc~wSHAcafjon Qr PQ~<~8<m t Srs z.c,.i.]5 g.Perform equ red local le rate tests, includi the primary conta nment air lock leakage rate testing in accordance vith the Primary Containment Leakage Rate Testing Program.Hote: An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.The acceptance criteria for air lock testing are: (1)Overall air lock leakage rate is g (0.05 La)vhen tested at g Pa.(2)For door seal leakage, the overall air lock leakage rate is g (0.02 La)vhen the air lock is pressurized to (g 2.5 psig for at least 15 minutes).sec WusA'C>cab'on 0)C~~6R di=nr g 7S g, g.i~BFR Unit 3 3.7/4.7-6 NBlDMST HO.2 03 pAGE 3 OF~ SFe C i Pica on p.(.I-2 h (1)at y ti it i dete ined hat th crit ion 4.A.2.g is exc ded, epair shal be in tiate immed atel 4euwcd l@hrn C.2.Aft'SC4 4Abn Co)(2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within hours following etection of cxcessivc local leakage, the reactor hall)trio v g repairs are i n o~3 effected and th OC Q I local leakage me the acceptance criterion as demonstrated by QZ retest.See 3iu+g;c~gn Foe ('-~gg)cg Qg, 8<<tsTs S.b.l.3<<~)K sc'cking The main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refueling outage.Zf the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall bc performed,to correct thc condition. BPR Unit 3 3.7/4.7-e PAGE~~"~

JUSTIFICATION FOR CHANGES BFN ISTS 3.6.1.2-PRIMARY CONTAINMENT AIR LOCK ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433.As a result the Technical Specifications should be more readily readable, and therefore,-understandable by plant operators as well as other users.The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting)is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change.A2 CTS 4.7.A.2.h(1) requires repairs to be initiated immediately when it is determined the criterion of 4.7.A.2.g is exceeded.CTS 4.7.A.2.g requires LLRTs to be performed in accordance with the Primary Containment Leakage Rate Testing Program (CTS 6.8.4.3).CTS 4.7.A.2.h(2) then allows 48 hours to demonstrate 4.7.A.2.g can be met following detection of excessive local leakage.Since repairs are typically initiated immediately and proposed Required Action C.1 for 3.6.1.2 requires action be initiated to evaluate the primary containment overall leakage rate using the current air lock results and ACTION A of ISTS 3.6.1.1 will only allow 1 hour to restore primary containment to OPERABLE status prior to requiring the initiation of a shutdown (reference Justification M2 for Specification 3.6.1.1), CTS 4.7.A.2.h(1) has been deleted.TECHNICAL CHANGES-MORE RESTRICTIVE The current requirements for the air lock are located within the primary containment TS requirements. The current definition of primary containment integrity requires only one air lock door to be closed and sealed (i.e., the seal mechanism intact and sealing the door).Thus, no actions are required if one door is inoperable provided the other door is OPERABLE, since primary containment integrity only requires the one door.The proposed LCO requires the entire air lock to be OPERABLE, BFN-UNITS 1, 2, L 3 Revision 0

JUSTIFICATION FOR CHANGES'FN ISTS 3.6.1.2-PRIHARY CONTAINHENT AIR LOCK which includes both doors, as well as the interlock mechanism and the leak-tightness of the barrel.ACTIONS are provided (proposed ACTIONS A and B)to ensure that if one door or its interlock mechanism is inoperable, the other door is closed, locked and periodically verified to be closed and locked.If the interlock mechanism is inoperable, an allowance is provided to open the door provided a dedicated individual controls the access.Notes are provided to allow, the locked closed'verification to be performed administratively if the door is in a limited access area.These two new actions are not applicable, however, if the entire air lock is inoperable (as stated in proposed Note 1 to both ACTIONS A and B).To ensure that the primary containment LCO will be entered if air lock leakage results in exceeding overall primary containment leakage, NOTE 2 to the ACTIONS is also included.Overall, these new ACTIONS provide additional restrictions to plant operation. CTS 4.7.A.2.h requires repairs to be initiated immediately when it is determined that the criterion of 4.7.A.2.g is exceeded and if conformance to these criterion is not demonstrated within 48 hours following detection of excessive local leakage, a reactor shutdown is required.ACTION C of the proposed Specification requires the licensee to initiate action to evaluate primary containment overall leakage rate using the current air lock test results immediately, verify an air lock door closed within 1 hour and restore the air lock to OPERABLE status within 24 hours.If required ACTION C and the associate Completion Time is not met, the unit must be in HODE 3 in 12 hours and HODE 4 in 36 hours.This is more restrictive than current requirements. This change adds a Surveillance to verify the interlock mechanism works properly (only one door can be opened at a time).This will ensure that one door is always closed which maintains containment integrity. The addition of new requirements represents a more restrictive change.The current requirements for the air lock are located within the primary containment TS requirements (CTS 3.7.A.2.a), which requires the primary containment to be OPERABLE at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the vessel.The proposed BFN ISTS 3.6.1.2 applicability is HODES 1, 2, and 3.This is more restrictive since CTS does not require the primary containment to be OPERABLE when in HODE 2, not critical, and<212'F.BFN-UNITS 1, 2,&3 Revision 0 P@GF g QF UNIT 3 CURRENT TECHNICAL SP ECIF ICATION MARKUP

4.7.A.a Co a 2oae m>Z.<l 3 R pp)scab'>I i b.C~Primary containment integrity shall be maintained at all times vhen thc reactor is critical or vhcn the reactor vater temperature is above 212 F and fuel ie in thc reactor veeec ing"open vessel" physics tests at pover levels not to exceed 5 HW(t).Primary containment integrity ie confirmed if the maximum allovablc integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.If E2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be rcduccd to<542 SCFH within 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.te ated Leak Rat est Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for thc last 24 hours.Excessive leakage ie indicated by a 82 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for drywell tcmperaturc, pressure, and venting operations) at 49.6 peig.Corrected to normal dryvcll operating pressure of 1.1 psigf this value is 542'SCFH.If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.Perform leakage rate testing in accordance with thc Primary Containmcnt Leakage Rate Testing Program.5<<WuSW Fs'cation Q<Q,Q>9'c BC'Sf'5 3,g,~,(BFH Unit 1 3.7/4.7-3 aMreMmr NO.228.PAGE~OF Cl~, 4.7.A.2.(Cont'd)Zf at any time it is determined that the criterion of 4.7.A.2,g is exceeded, repairs shall be initiated immediately. +~'e+~kÃi~4it'Gr 6 Aptly 4~gf'N IsrS g.g/.Iyg~(p (2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours f ollowing detection of excessive local leakage, the reactor shall be shut down and depressurixed until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.sg s.6./.3./o tK isolation valves shall be tested at a pressure of 25 psi for leakage durin ach refuels.n p2.outa e Zf t e leakage rate of 1l.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 1 3.7/4.7-8

3.7.C 3-Secondary containmcnt integ-rity shall be maintained in the refueling xone, except as specified in 3.7.C.4.4.If refueling zone secondary containment cannot bc maintained the following conditions shall be mct: a.Handling of spent fuel and all operations over spent fuel pools and open reactor~elis containing fuel shall be prohibited. b.The standby gas treatment system suction to thc refueling zone vill be blocked~cept for a controlled leakage area sized to assure the achieving of a vacuum of at least 1/4-inch of water and not over 3 inches of~ater in all three reactor zones.This is only appli-cablc if reactor xone integrity is required.ey.cgt e Keac~c Suig g lfd culm br<ag~g~36./.3 Mhen Primary Containment Integrity is required, all rimary containment isolati valves and all reactor coolant system instrument line floe check valves s 11 bc OPERhB except as specified in 3.7.D.2.*Locked or sea cd closed valves may be opened on an inter-mittent basis under administrative control.a.~<3,e.l,3.5 CR p.g,!,~.to SP.g,v, l.3.1 At least once r o cr-ating c clc the OPER-primary contain-ment isolation valves that are po~er operated and automatically initiated shall be tested for simulated automatic initiation Ac os.Ll 1.The primary containment isolation valves surveillance s6all bc performed as follows: SFN Unit 1 3.7/4.7-17 SlegMNr N0.y 8 g

MckEI~Qai SR 3.i.1.3.5 az,c,.l z d ia accordance vith Specification lo0ol%lg tested for closure times.f<8'M HC1(op)8 t'r.e4ug 8<T>oz p (Q r sid kCr i~)F b.In accordance with Specification 1.0.%5, all normally open povcr operated primary containment isolation valves shall bc fuactionally tested.f~~s~4 AJ4~onl5 1<4 used ALIAS 3+9 H Q.TiOWX C~~<3.4.1.'3 t d EF'CV ac l4 Bc iSo)pHo PsiW~o n Sindhi tah J i tlstvu~g+/q Wcc t signai (Deleted)At leas once pcr operatin c cl the OP ILITY of the reactor coolant system instrument line floe check valves shall be verified.QTlo PyG!2.Ia the even any primary c tai olation valve becomes inoperable, reactor operation may continue provided l~nl'how at least oae valve, ia each line having an inoperable valve, is OPERhBLE and s eithers Lq 20<<Owu<A ItC Tip P~r 4t.'o~Mhenever a primary contaiamcnt isolation valve is inoperable, thc position of at least the valv each line having an ino rable valve shall be ecor e daily a.The inoperablc valve is restored to OPERABLE status, or b.Each affected line is.isolated by use of at least one deactivated contaiameat RC7(od isolation valve secured ia the isolated position.3.If Specification 3.7.D.l and 3 7.D.2 cannot be mct, aa orderly shutdcnm shall be~<<4" initiated and the reactor shall g bc ia th COLD SHUTDOWN CONDITION+it hour's~l(rtosedPs P,t,l, Z.)3o 0 el i3.Z 3.4~l.3.'3 3as ol g,s,, ls 3,g<loSed mang~l Valve>gl;+~l~<~o<<hect Vole W4~Cl~4h~~gh<he Valge Sec~rg L3 SpÃUnit 1 He HoTSNutnowu C~on->o~<n iZ+~vs@gal.iA 3.7/4.7-18 NENOMHfT N5.f 8 9 i ACi"~C-

5ftC<4<rabO 3.t..l.3 FEB 13 19%3.7.F.4.7.F..The primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3.7.F.2.a.The results of the ia-place cold DOP and hslogenatcd hydrocarbon tests at design flovs on HEPA filters and charcoal adsorber banks shall shov g 99K DOP removal and g 99K halogcnated hydro-carbon removal vhen tested in accordance vith AHSI H510-1975. b.The results of laboratory carbon sample analysis shall shov g 85K radioactive methyl iodide removal vhca tested ia accordance vith ASTN D3803.c.System flov rate shall bc shown to be vithin g lOX of design flov vhea tested ia accordance with ASSI H510-1975'e ao The 18-inch primary contain-acnt isolation valves asso-ciated vith PURQIHC may be open during the RUN mode for a 24-hour period after enteriag the RUN aodc aad/or for a 24-hour period prior to entering the SHUTDOWH mode.e OPERABILITX o BPS Unit 1 c.co Z 6.i.~3.7/4.7-21 2.If the provisions of 3.7.F.l.a, b, and c cannot be aet, the system shall be declared iaoyerable. The provisions of Technical Spccificatioa 1.C.l do aot apply.PURCIHC aay coa-thrns using the Staadby Qas Treatment Systea l.At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorbcr banks shall be demonstrated to be less than 8.5 inches of vater at system design flow rate{g 10%).a.The tests and sample analysis of Specifica-tion 3.7.F.1 shall be performed at least once per operatiag cycle or once every 18 months, vhichever occurs first or after 720 hours of system operation and following significant painting, fire, or chemical release in any ventilation zone commmicating vith the systcao b.Cold DOP testing shall be performed after each complete or partial replacement of the HEP filter bink or after any structural mainte-nance on the system housing.c.Halogenatcd hydrocarboa testing shall be performed after each complete or partial reylacement of the charcoal adsorber bank or after any structural aaintenancc on the system housing.S<.'e 5<bagl~Hon Q Qgg<g 4'r Cy5 37 F/q<7~F ln t.h4'c<hon g, (,].3 APA 29 tg9t these primary~'~'i'~containment fsolation valves is governed by Technical Specification 3.7.D.b.Pressure control of the cont nment i norma ly perfo ed by IHQ through 2-fnch rfmary ontainm t isol tion ives vh ch rout ef luent t the St dby Gas reatmen System.The EIUNILI o f these rimary contaf ent iso tion valves i governe by Technical pecification 3.7.D.Z.A (3.7.G.1.The Containment Atmosphere Dilution (CiD)System shall be OPERABLE vith: a.Tvo independent systems capable of supplying nftrogen to the dryvell and torus'.Cycle each solenoid operated air/nitrogen valve through at least one complete cycle of full travel fn accordance vfth Specification 1.0.MK, and at least once per month verify that each manual valve fn the flov path is open.BPS th6t 1 b.A minimum supply of 2,500 gallons of lfqafd nitrogen per systems<<St'.t-I C+O~Q(C~4<4".7/4.7-22 b.Veri that the CAD System contains a minimum supply of 2,500 gallons of lfgafd nitrogen tvic er veek.ENDMae NtL Z se PAGE~OF~ INSERT PROPOSED NEW SPECIFICATION 3.6.1.4 Insert new Specification 3.6.1.4,"Drywell Air Temperature," as shown in the BFH Unit 2 Improved Standard Technical Specifications. 0 JUSTIFICATION FOR CHANGES BFN ISTS: 3.6.1.4-DRYMELL AIR TEHPERATURE TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new Specification is being added requiring drywell air temperature to be 150'F.This is required since some accident analyses assum'e this temperature at the start of an accident.Appropriate ACTIONS and Surveillance Requirements are also added.This is consistent.with the BWR Standard Technical Specifications, NUREG 1433.0 BFN-UNITS 1, 2, 8L 3 Revision 0.u 0 UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP PAGE OF 5'eCe.hCa Ond b f Al 4 7 A.a Co 2's LCo 3.a.l.3 PPPli Cab~t e b.C~Primary containment integrity shall be maintained at all times vhen the reactor is critical or.vhen the reactor vater temperature is above 212 F and fue s in the reacto esse ccpt w c performing"open vessel" physics tests at power levels not to exceed 5 Kf(t).Primary containment iategrity is confirmed if the maximum allovable integrated leakage rate, La, does not exceed the equivalent of 2 percent of thc primary containment volume pcr 24 hours at the 49.6 psig design basis accident pressure, Pa.If 52 makeup to the primary containment averaged over 24 hours (corrcctcd for pressure, temperatures aad venting operations) exceeds 542 SCFH, it must bc reduced to<542 SCFH vithin 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.e ae Primary containment nitrogen consumption shall bc monitored to determine the average daily'nitrogen consumption for thc last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary contaiamcnt free volume per 24 hours (corrcctcd for dryvell temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal dryvell operating prcssure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified ia 3.7.h.2.C shall be taken.Perform leakage rate testing ia accordance vith thc Primary Containmeat Leakage Rate Tcstiag Program.5ee&~we'eeei n CeeChuys p BF<isis 3 r.i.l~0 BPS Unit 2 3.7/4.7-3 hMENStEÃf NL 2 4 3 FAGE~OF 7 +ciCAfiow s.C./.3 4.7.A.4.7.A.2.(Cont'd)(1)If at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated immediately. ~c>$446$io~p C%~5'~gfh/IS7S 3.6.I./+3.C./.2.(2)If conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.3.C./.'R The main steamlxne isolation valves shall be tested at a pressure of 25 psig for leakage during eac re ue x If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 2 3.7/4.7-8 PAGE

3.7,C, Se o d Co ta ent 3.Secondary containment integ-rity shall be maintained in the refueling zone, except as specified in 3.7.C.4.qgP VASTl AJAR'iON FOR gpwN~s Fog ZHiv ad 3.6.V./4.If refueling zone secondary containment cannot be maintained the folloving conditions shall be met: a.Handling of spent fuel and all operations over spent fuel pools and open reactor wells containing fuel shall be prohibited. b.The standby gas treatment system suction to the refueling zone vill be blocked except for a controlled leakage area sized to assure the achieving of a vacuum of at least 1/4-inch of vater and not over 3 inches of eater in all three reactor zones.This is only appli-cable if reactor zone integrity is re uired.ms+paa J r 1b'll.l, QC,(aalesara Qrteakg r(ima V e oato ma V e tai t Isol on 1.The primary containment isolation valves surveillance shall be performed as follows: Q~Shen Primary Containment Integrity is required, all rimary containment isolation valves and all reactor coolant system instrumen line flow check valves shall be OPERAB except as specified in 3.7.D.2.~.l.g~li~ob.IA LC.O 3.lo./, 3 a.At least once per o er-ti c c the OPER-ABLE primary contain-ment isolation valves that are pover operated and automatically initiated shall be tested for imulated automatic initiation act+ad.or I NENOMEgr go.2 0 g sR 3.4.1.3.5 SR~4.I.M SR3.l'm.1.3.7 go+(~+CTIONg<gag ko SR.3.C.t)3.7/4.7-17 BFH Unit 2 PAGE t OF I*Locked or sealed closed valves may be opened on an intermittent basis under administrative control. ' Vaa'yes at L5 P~P~AC7.(oe 8 Propos 4'o6l D sR 3.g,/,3p SR~C.(.3.g and in accordance with Specification 1.0.MM, tested for closure times.In accordance with Specification 1.0.MN, all normally open power operated primary containment isolation valves shall be functionally tested.PmposM hcT(oQ p~"o~c'4 No>4.2 4a ACAahP/6 Propos~(" cs g l 4ci c7cogg c.(Deleted)Sg 3.C.l.8.8 d.tFCV w+Q 4~iso/aA'm p4S>)i~o>f s:~/~~lh jfrlf~<g/l~C$~S)vo At least nce pe o erati'c c the OPERABILITY of the reactor coolant system instrument line flow check valves shall be verified.A('TlOhLS P,e+2~CoA ki Ae c.2.~A/4<<cg A ck~4.Z+C,2.enever a primary contain-ment isolation valve is inoperable, the osition of t least one other valve in each line having an ino erable valve shall be recor e ail a.The inoperable valve is restored to OPERABLE status, or~roPoL~Sls 3 Co.I 3 3.C, L3.2S.C.J 3.3 3.4/2.g Z.C.(P.g b.gcyuirQ Aa4ia~.R.I+C.[Each affected line is isolated by use of at least one deactivated containment isolation valve secured in the isolated positio mc wag w~Ip4 f,/;g"k i o~+Ice(c.vg/P/~+J~lr~/~Secor~If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be i t COLD SHUTDOWN COHDITIOH within ours.3.AC.<lou E, AMENDMEHT NO 2 0 4 5 o;,V 3.7/4.7-18 3&BFH La Unit 2 Pl 8~'+ggacVlA~~~ogo iTic (2 lip,~In the event any pr mary conta n ent o va v becomes inoperable, reactor operation may continue provided at least one valve, in each line having an inoperable valve, is OPERABLE and thin 4 hours either: LY ~7.F.o t 4.7.F.~Ss~te 0 1.Thc primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3..7.F.2.a.The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show g 99K DOP removal and g 99K halogenated hydro-carbon removal vhen tested in accordance vith AHSI H510-1975. b.The results of laboratory carbon sample analysis shall shov g 85K, radioactive methyl iodide removal vhcn tested in accordance vith ASTI D3803.c.System flov rate shall be shown to be vithin g 1OX of design flov vhcn tested in accordance vith AHSI H510-1975~2.If the provisions of 3.7.F.l.a, b, and c cannot be met, the system shall be declared inoperable. The provisions of Technical Specification 1.C.l do not apply.PURGIHG may con-tixxae using the Standby Gas Treatment System.3o ae SR s.<1.3.I f4of~The 18-inch primary contain-ment isolation valves asso-ciated vith PURGIHG may be open during the RUH mode for a 24-hour period after entering the RUN mode and/or for a 24-hour period.prior to eater the SHUTDOMN mode.e 0 TY of BPK-~-'~->"....3.7/4.7-21 Unit 2-1.At least once every 18 months, the prcssure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 8.5 inches of vater at system design flov rate (g lOX).a.The tests and sample analysis of Specifica-tion 3.7.F.1 shall be performed at least once per operating cycle or once every 18 months, vhichever occurs first.or after 720 hoars of system operation and f olloving significant painting, fire, or chemical release in any ventilation xone communicating vith thc systems b.Cold DOP testing shall be performed after each complete or partial replacement of the HEPA~filter bank or after any structural maintc nance on the system housing.c.Halogeaatcd hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.SEE&$TI Pic'&ion)f og<<""-5$F~~c~S g 7~/q Sec4io~NENMRf NO 231

APR 2 9 1991~s.a e these primary/Co 9.C.t.3 containment isolation valves is governed by Technical Specification 3.7.D.b.Pressure control of the contai nt is normally performed by VENTING through 2-ch primary containment olation valves vhich ute ffluent to the Standby G Treatment Sy em.The PERABILITY o these rimary contai ent isolatio valves governed by Technical ecification 3;7.D.3.7.G.Co ta e mos ere ut o S ste C 1.The Containment Atmosphere Dilution (CAD)System shall be OPERABLE vith: 4.7.G.Co ta e t t os ere utio S stem C D 1.S st 0 crab t a.Tvo independent systems capable of supplying nitrogen to the dryvell and torus.a.Cycle each solenoid operated air/nitrogen valve through at least one complete cycle of full travel in accordance vith Specification 1.0.MM, and at least once per month verify that each manual valve in the flov path is open.~..Unit 2 b.A minimum supply of 2,500 gallons of liquid nitrogen per system..7/4.7-22 5 E~0<S TI F'C AT t o g F'o~C9+n'~Fag.Pp/J<<~Z.g.g,i~b.Verify that the CAD System contains a minimum supply of 2,500 gallons of liquid nitrogen tvice per veek.AMENDMEMT N0.I 9 7'AG~~o~~ 0' INSERT PROPOSED NEW SPECIFICATION 3.6.1.4 Insert new Specification 3.6.1.4,"Drywell Air Temperature," as shown-in the BFN Unit 2 Improved Standard Technical Specifications. JUSTIFICATION FOR CHANGES BFN ISTS: 3.6.1.4-DRYWELL AIR TEMPERATURE TECHNICAL CHANGES-NORE RESTRICTIVE Hl A new Specification is being added requiring drywell air temperature to be 150'F.This is required since some accident analyses assume this temperature at the start of an accident.Appropriate ACTIONS and Surveillance Requirements are also added.This is consistent.with the BWR Standard Technical Specifications, NUREG 1433.BFN-UNITS 1, 2, 5, 3 Revision 0

UNIT 3 CURRENT TECHNICAL SPECIFICATION 2oae+l>CA bo l e'fy b.co Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water t'emperature is above 212 F and fuel is in the reactor vessel e ing"open vessel" physics tests at power levels not to exceed 5 MW(t).Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours at the 49.6 psig design basis accident pressure, Pa.If H2 makeup to the primary containment averaged over 24 hours (corrected for pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced to<542 SCFH within 8 hours or the reactor shall be placed in Hot Shutdown within the next 16 hours.2.te rat d e ate est Primary containment nitrogen consumption shall be monitored to determine the average daily.nitrogen consumption for the last 24 hours.Excessive leakage is indicated by a H2 consumption rate of>2X of the primary containment free volume per 24 hours (corrected for drywall temperature, pressure, and venting operations) at 49.6 psig.Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH.If this value is exceeded, the action specified in 3.7.k.2.c shall be taken.Perform leakage rate testing in accordance with the Prima Containment Leakage Rate Testing Program.e g~hknlfio~ kR Qagr5 foA BFQ[5+5 g,4g,/BFK Unit 3 3.7/4.7-3 NENMERr go.2 Og ma~~

4.7.A.4.7.A.2.(Cont'd)h.(1)Zf at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated immediately. ~<~+~$5~~4%40~4" C(~~Ar it/-nl/st y.g,/.I~~~I+(2)Zf conformance to the criterion of 4.7.A.2.g is not demonstrated within 48 hours following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.SR Z.C./3.~o main steamline isolation valves shall be tested at a pressure of 25 psig for leakage during each refuelzn outa e Zf the leakage rate of 11.5 scf/hr f or any one main steamline isolation valve is exceeded, repairs and retest shall be performed to correct the condition. BFN Unit 3 3.7/4.7-8 0 3.7.C 3.5ccondary contaiameat integ-rity shall bc maintained in the rcfueliag zone, except as specified in 3.7.C.4.4.Ig refueliag zone secondary containment cannot be maintained the folloving conditions shall bc met: cc wushCi'ccgi~ Ch~~, 8c'N isfs w.~.q,j a.Handliag of spent fuel and all operations over spent fuel pools and open reactor vclls coataining fuel shall bc prohibited. The standby gas treatment system suction to the refucliag xone vill bc blocked except for a controlled leakage area sized to assure the achieviag of a vacuum of at least 1/4-inch of vater aad not over 3 inches of vater in all three reactor zones.This is only appli-cable if reactor zone integrity is required.epee~geaH+v pcgguW+<at'c<<A pcVion F b.In accordance vith Specification 1.0.MN, all normally open pover operated primary containment isolation valves shall be'unctionally tested P cnuz A+~2.eoAcii lion A+e.PapoScd Noh W 4 Rch'on5~oP<Srd N&S 3+/+AonS In the even primary contain ent isolation valv ccomes o c, reactor operation may continue provided at least one valve, in each line having an inoperablc valve, is OPERABLE and our either: The inoperable valve is restored to OPBRABLB status, or%H n 4,+C inoperable, the position of at least one valv in each line having an ino erablc valve shall be ordcd da y.f Oops'tl SRs 3 c.(Deleted)5 Z d.At least once per operating cycl the ECe4 a Ww oPB o the fo%t<5ol atia&reactor coolant system fes:how on a instrument linc flov check valves shall be verified linc ggav.S ignis 2.whenever a primary contain-ment isolation valve is b EcQa6rclk 4@+n g,l+4al Each affected linc is isolated by use of at least one deactivated containment isolation valve secured the isolated position.3.If Specification 3.7.D.l and 3.7.D.2 cannot bc mct, an t orderly shutdovn shall bc initiated and the reactor shall G bc th COLD SHUTDOWN(COHDITIOK vi hours'i BPK 3.7/4.Unit 3<AT>Halo~Coa)Dido&LR~s 4&l~AI 7-18 L3~4.I.S.3 3'o c,l,g,g.l.3.$AMENDMENT NO.I 6 y PAGE<losed~n~)Value,gl;nd ~<a%<,~caw.i va>n~,~ohe vaNe gecur~ot Q'e cia'I~3 FEB 1 81995.7.F.t c 4.7.F.~Ssteg 1.The primary containment purge system shall be OPERABLE for PURGIHG, except as specified in 3.7.F.2.a.The results of thc in-place cold DOP and halogenated hydrocarbon tests at design flovs on HEPA filters and charcoal adsorber banks.shall shov g 99Z DOP removal and g 99Z halogenated hydro-carbon removal vhcn tested in accordance vith AHSI H510-1975. k b.The results of laboratory carbon sample analysis shall shov g 85Z radioactive methyl iodide removal vhcn tested in accordance vith AS'3803~c.System flov rate shall bc shovn to be vithin g 10Z of design flov vhcn tcstcd in accordance vith AHSI H510-1975~3o ao SC]]