ML23275A111
| ML23275A111 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/02/2023 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23275A109 | List: |
| References | |
| L-2023-126 | |
| Download: ML23275A111 (1) | |
Text
St. Lucie Plant Units 1 and 2 L-2023-126 Dockets 50-335 and 50-389 ENCLOSURE 2 ST. LUCIE PLANT, UNITS 1 AND 2 IMPROVED TECHNICAL SPECIFICATIONS (ITS) REVISION 2 SUBMITTAL VOLUMES 1 THROUGH 16 (4915 TOTAL PAGES, INCLUDING COVER SHEETS)
ENCLOSURE 2 VOLUME 1 ST. LUCIE PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION APPLICATION OF SELECTION CRITERIA TO THE ST. LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Revision 2 R2
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Revision 2 R2
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS i
CONTENTS Page
- 1.
INTRODUCTION....................................................................................................... 1
- 2.
SELECTION CRITERIA............................................................................................. 2
- 3.
PRA INSIGHTS......................................................................................................... 5
- 4.
RESULTS OF APPLICATION OF SELECTION CRITERIA........................................ 8
- 5.
REFERENCES......................................................................................................... 9 ATTACHMENT
- 1.
SUMMARY
DISPOSITION MATRIX FOR ST LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 APPENDIX A.
JUSTIFICATION FOR SPECIFICATION RELOCATION
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 1 of 9
- 1.
INTRODUCTION The purpose of this document is to confirm the results of the Combustion Engineering (C-E)
Owners Group application of the Technical Specification selection criteria on a plant specific basis for the St. Lucie Nuclear Plant (PSL) Unit 1 and Unit 2. Florida Power & Light Company (hereinafter FPL) has reviewed the application and confirmed the applicability of the selection criteria to each of the Technical Specifications utilized in report CEN-355, "C-E Owners Group Restructured Standard Technical Specifications Volume 1 Criteria Application," (Reference 1),
NRC Staff Review of nuclear steam supply system (NSSS) Vendor Owners Groups Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications, Wilgus/Murley letter dated May 9, 1988 (Reference 2), and as revised in NUREG-1432, Revision 5.0 "Standard Technical Specifications, Combustion Engineering Plants" (Reference 3) and applied the criteria to each of the current PSL Technical Specifications. Additionally, in accordance with the NRC Final Policy Statement (Reference 4), this confirmation of the application of selection criteria includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in Reference 1, as applicable to the PSL.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 2 of 9
- 2.
SELECTION CRITERIA FPL has utilized the selection criteria provided in Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36), dated July 19, 1995 (Reference 5), which were codified following issuance of the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (Reference 4) to develop the results contained in the attached matrix. PRA insights as used in the C-E Owners Group submittal were utilized, confirmed by FPL, and are discussed in the next section of this report.
The selection criteria and discussion provided in Reference 4 are as follows:
Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary:
Discussion of Criterion 1: A basic concept in the adequate protection of the public health and safety is the prevention of accidents. Instrumentation is installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safely, thus reducing the likelihood of a loss-of-coolant accident.
This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage. This criterion should not, however, be interpreted to include instrumentation to detect precursors to reactor coolant pressure boundary leakage or instrumentation to identify the source of actual leakage (e.g.,
loose parts monitor, seismic instrumentation, valve position indicators).
Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier:
Discussion of Criterion 2: Another basic concept in the adequate protection of the public health and safety is that the plant shall be operated within the bounds of the initial conditions assumed in the existing 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
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 3 of 9 the Final Safety Analysis Report (FSAR), for which a structure, system, or component must meet specified functional goals. These analyses are contained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as Condition II, III, or IV events (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 and transient analyses even if they cannot be directly observed in the control room (e.g, moderator temperature coefficient and hot channel factors).
The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA and transient analyses, and which are monitored and controlled during power operation. If these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure/low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.
Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a 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
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 4 of 9 recent years and provide a method of presenting the plant response to an accident. These can be used to define the primary success paths.
A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's DBA and transient analyses, as presented in Chapters 6 and 15 of the plant's Final Safety Analysis Report (or equivalent chapters). Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to 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 policy that pressurized water reactor (PWR) licensees retain in their Technical Specifications LCOs, action statements and Surveillance Requirements for the Residual Heat Removal System, which operating experience and PRA have generally shown to be significant to public health and safety, and any other structures, systems, or components that meet this criterion.
The Commission recognizes that other structures, systems, or components may meet this criterion. Plant and design-specific PRA's have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report DBA or transient analyses. It is
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 5 of 9 the intent of this criterion that those requirements that PRA or operating experience exposes as significant to public health and safety, consistent with the Commission's Safety Goal and Severe Accident Policies, be retained or included in Technical Specifications.
The Commission expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant specific PRA or risk survey and any available literature on risk insights and PRAs. This material should be employed to strengthen the technical bases for those requirements that remain in Technical Specifications, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.
Similarly, the NRC staff will also employ risk insights and PRAs in evaluating Technical Specifications related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.
- 3.
PRA INSIGHTS Introduction and Objectives Reference 4 includes a statement that NRC expects licensees to utilize any plant specific PRA or risk survey and any available literature on risk insights and PRAs to strengthen the technical bases for these requirements that remain in Technical Specifications and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.
Those Technical Specifications proposed as being relocated to other plant controlled documents will be maintained under programs subject to the 10 CFR 50.59 review process. These Relocated Specifications have been compared to a variety of PRA material with two purposes:
- 1) to identify if a Specification component or topic is addressed by PRA; and 2) if addressed, to judge if the Relocated Specification component or topic is risk-important. The intent of the PRA review was to provide an additional screen to the deterministic criteria. This review was accomplished in the generic C-E Owners Group submittal CEN-355 (Reference 1). The results
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 6 of 9 of this generic review have been confirmed by FPL for the applicable PSL Specifications to be relocated.
Assumptions and Approach The results of the generic criteria application documented in CEN-355 were reviewed to identify systems and components to be relocated from the Technical Specifications to other means of control. This list was compared to the findings from several PRAs to determine if any of the items proposed to be removed from the Technical Specifications were risk significant.
At the time CEN-355 was issued, the only published PRA for a C-E designed NSSS was the Calvert Cliffs Unit 1 Interim Reliability Evaluation Program (IREP) (NUREG/CR-3511, Interim Reliability Evaluation Program: Analysis of the Calvert Cliffs Unit 1 Nuclear Power Plant," dated August 1984). This IREP was examined to identify the systems and components in the risk dominant event sequences. Event sequences with lower frequency of occurrence than the cutoff of 1.0 E-6 per year were also reviewed to ensure that systems of prime importance were not overlooked. Tables 3-1 and 3-2 of CEN-355 summarized the results of the examination of the risk significant event sequences. As noted in CEN-355, Calvert Cliffs Unit 1 is of the same design as PSL Units 1 and 2.
Table 3-1 represented those systems and components found to be risk significant; i.e., systems and components required to terminate or mitigate the event sequences leading to core melt having a frequency of occurrence of at least 1.0 E-6 per year.
Table 3-2 identified the dominant accident sequences identified as leading to core melt which have a frequency greater than E-6 per year.
To ensure an adequate representation of plants with a C-E designed NSSS, several partial PRAs for plants with a C-E NSSS were examined. These included:
CEN-239, Probabilistic Risk Assessment of the Effect of PORVs on Depressurization and Decay Heat Removal, Supplements 1, 2, and 3, June 1983; PRA evaluations of Palisades main steam line break events; and
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 7 of 9 PSL Shutdown Decay Heat Removal Analysis NUREG/CR-4710, Shutdown Decay Heat Removal Analysis, Combustion Engineering 2-Loop PWR Case Study, 1987.
Also, full PRAs on PWRs with other NSSS designs were examined. These included:
Oconee, a two loop Babcock and Wilcox PWR, and Seabrook, a four loop Westinghouse PWR.
To ensure that no risk-significant systems were overlooked, a listing of the requirements not meeting criteria for retention as technical specifications was reviewed and documented in Table 3-3 of CEN-355, Appendix B.
Appendix C of CEN-355 includes safety sequence diagrams for identified types of initiating events considered in formal risk assessments. Table C 1 in Appendix C, CEN-355 tabulated the events for which the safety sequence analysis (SSA) were performed. The results of the SSA are summarized in the sequence of events diagrams provided in Table C.1 of Appendix C, CEN-355.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 8 of 9
- 4.
RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 herein were applied to the PSL Unit 1 and Unit 2 Technical Specifications. The following Summary Disposition Matrix for St. Lucie Unit 1 and Unit 2 is a summary of that application indicating which Specifications are being retained or relocated, the criteria for inclusion, if applicable, the NRC results of the criteria application as expressed in the NRC Staff Review of NSSS Vendor Owners Groups Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications, Wilgus/Murley letter dated May 9, 1988 (Reference 2), and any necessary explanatory notes. Discussions that document the rationale for the relocation of each Specification which failed to meet the selection criteria are provided in Appendix A, except as noted in the Summary Disposition Matrix.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Page 9 of 9
- 5.
REFERENCES
- 1.
CEN-355, "C-E Owners Group Restructured Standard Technical Specifications Volume 1 Criteria Application, December 1987.
- 2.
Letter from T.E. Murley (NRC) to W.S. Wilgus (B&W Owners Group), NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups Application Of The Commissions Interim Policy Statement Criteria To Standard Technical Specifications, dated May 9, 1988.
- 3.
NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants,"
Revision 5.0, September 2021.
- 4.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (Federal Register Notice 58 FR 39132).
- 5.
10 CFR 50.36, Technical specifications, (c)(2)(ii) selection criteria, July 19, 1995 (Federal Register Notice 60 FR 36953).
ATTACHMENT 1
SUMMARY
DISPOSITION MATRIX FOR ST LUCIE UNIT 1 AND UNIT 2
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 1 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 1.0 DEFINITIONS 1.1 Yes This section provides definitions for several defined terms used throughout the remainder of Technical Specifications. They are provided to improve the meaning of certain terms.
As such, direct application of the Technical Specification selection criteria is not appropriate. However, only those definitions for defined terms that remain because of application of the selection criteria, will remain as definitions in this section of Technical Specifications.
2.1 SAFETY LIMITS 2.1.1 Reactor Core 2.1.1 Yes Application of Technical Specification selection criteria is not appropriate.
However, Safety Limits will be included in Technical Specifications as required by 10 CFR 50.36.
2.1.2 Reactor Coolant System Pressure 2.1.2 Yes Same as above.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 2 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 Reactor Trip Setpoints except Unit 2 Table 2.2-1, Functional Unit 10 3.3.1 Yes-3 The Reactor Protection System (RPS) limited safety system settings (LSSS) have been included as part of the RPS instrumentation Specification, which has been retained since the Functions either actuate to mitigate consequences of design basis accidents and transients or are retained as directed by the NRC as the Functions are part of the RPS.
NA 2.2.1 Reactor Protective Instrumentation Trip Setpoint Limits Unit 2 Table 2.2-1, Functional Unit 10 Relocated No See Appendix A, page 1.
3/4.0 APPLICABILITY 3.0.1 Limiting Condition For Operation LCO 3.0.1 Yes This Specification provides generic guidance applicable to one or more Specifications. The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1432, Revision
- 5.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 3 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3.0.2 Limiting Condition For Operation LCO 3.0.2 Yes Same as above.
3.0.3 Limiting Condition For Operation LCO 3.0.3 Yes Same as above.
3.0.4 Limiting Condition For Operation LCO 3.0.4 Yes Same as above.
3.0.5 Limiting Condition For Operation LCO 3.0.5 Yes Same as above.
3.0.6 Limiting Condition For Operation LCO 3.0.6 Yes Same as above.
4.0.1 Surveillance Requirements SR 3.0.1 Yes Same as above.
4.0.2 Surveillance Requirements SR 3.0.2 Yes Same as above.
4.0.3 Surveillance Requirements SR 3.0.3 Yes Same as above.
4.0.4 Surveillance Requirements SR 3.0.4 Yes Same as above.
4.0.5 Surveillance Requirements Deleted No Deleted. See CTS 4.0.5 technical change discussion in the Discussion of Changes for ITS Section 3.0.
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 4 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.1.1.1 Shutdown Margin - Tavg > 200°F 3.1.2 3.1.4 3.1.5 3.1.6 Yes-2 3/4.1.1.2 Shutdown Margin - Tavg 200°F 3.1.1 Yes-2 3/4.1.1.3 Boron Dilution Relocated No See Appendix A, page 3.
3/4.1.1.4 Moderator Temperature Coefficient 3.1.3 Yes-2 3/4.1.1.5 Minimum Temperature for Criticality 3.4.2 Yes-2 3/4.1.2 BORATION SYSTEMS 3/4.1.2.1 Flow Paths - Shutdown Relocated No See Appendix A, page 4.
3/4.1.2.2 Flow Paths - Operating Relocated No See Appendix A, page 4.
3/4.1.2.3 Charging Pumps - Shutdown Relocated No See Appendix A, page 4.
3/4.1.2.4 Charging Pumps - Operating Relocated No See Appendix A, page 4.
3/4.1.2.5 Boric Acid Pumps - Shutdown Relocated No See Appendix A, page 4.
3/4.1.2.6 Boric Acid Pumps - Operating Relocated No See Appendix A, page 4.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 5 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.1.2.7 Borated Water Sources - Shutdown Relocated No See Appendix A, page 4.
3/4.1.2.8 Borated Water Sources - Operating Relocated No See Appendix A, page 4.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES 3/4.1.3.1
[Full Length - Unit 1] CEA Position 3.1.4 Yes-2, 3 3/4.1.3.3 3/4.1.3.2 Position Indicator Channels [- Operating -
Unit 2]
3.1.7 Yes-2 Based on NUREG-1431 (Westinghouse STS) ISTS 3.1.7 NA 3/4.1.3.3 Unit 2 only - Position Indicator Channels -
Shutdown Relocated No See Appendix A, page 6 3/4.1.3.4 CEA Drop Time 3.1.4 Yes-2 3/4.1.3.5 Shutdown CEA Insertion Limit 3.1.5 Yes-2 3/4.1.3.6 Regulating CEA Insertion Limits 3.1.6 Yes-2 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Linear Heat Rate 3.2.1 Yes-2 3/4.2.3 Total Integrated Radial Peaking Factor - FTr 3.2.2 Yes-2 ISTS 3.2.3 3/4.2.4 Azimuth Power Tilt - Tq 3.2.3 Yes-2 ISTS 3.2.4
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 6 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.2.5 DNB Parameters 3.2.4 3.4.1 Yes-2 ISTS 3.2.5 - ASI ISTS 3.4.1 - RCS Flow Rate, RCS Cold Leg Temp, Pressurizer Pressure 3/4.3 INSTRUMENTATION 3/4.3.1 Reactor Protective Instrumentation except Unit 2 Table 3.3-1, Functional Unit 10 3.3.1 3.3.2 3.3.11 3.3.12 Yes-3 ISTS 3.3.1 ISTS 3.3.3 ISTS 3.3.13 ISTS 3.3.2 NA 3/4.3.1 Reactor Protective Instrumentation Unit 2 Table 3.3-1, Functional Unit 10 Relocated No See Appendix A, page 1 3/4.3.2 Engineered Safety Feature Actuation System Instrumentation 3.3.3 3.3.4 3.3.5 Yes-3 ISTS 3.3.4 ISTS 3.3.5 ISTS 3.3.6 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 Radiation Monitoring 3.3.6 3.3.7 3.3.8 3.4.15 Yes-3 ISTS 3.3.7 ISTS 3.3.8 ISTS 3.3.10 ISTS 3.4.15 3/4.3.3.5 Remote Shutdown [Instrumentation - Unit 1]
[System Instrumentation - Unit 2]
3.3.10 Yes-4 ISTS 3.3.12 R2
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 7 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.3.3.8 3/4.3.3.6 Accident Monitoring Instrumentation (Type A and non-Type A Category 1 Instruments) 3.3.9 Yes-3, 4 ISTS 3.3.11 3/4.3.3.8 3/4.3.3.6 Accident Monitoring Instrumentation (Non-Type A and not Category 1 Instruments) 3.3.9 No See Appendix A, page 7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.1.1 Startup and Power Operation 3.4.4 Yes-2, 3 3/4.4.1.2 Hot Standby 3.4.5 Yes-3 3/4.4.1.3 Hot Shutdown 3.4.6 Yes-4 3/4.4.1.4.1 Cold Shutdown (Loops Filled) 3.4.7 Yes-4 3/4.4.1.4.2 Cold Shutdown (Loops Not Filled) 3.4.8 Yes-4 3/4.4.3 3/4.4.2.2
[Safety Valves - Operating - Unit 1]
[Operating - Unit 2]
3.4.10 Yes-3 3/4.4.4 3/4.4.3 Pressurizer 3.4.9 Yes-2, 3 3/4.4.5 Steam Generator (SG) Tube Integrity 3.4.17 Yes-2 ISTS 3.4.18
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 8 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakage Detection Systems 3.4.15 Yes-1 3/4.4.6.2
[Reactor Coolant System - Unit 1]
[Operational - Unit 2] Leakage 3.4.13 3.4.14 Yes-2 3/4.4.8 Specific Activity 3.4.16 Yes-2 3/4.4.9 PRESSURE/TEMPERATURE LIMITS 3/4.4.9.1 Reactor Coolant System 3.4.3 Yes-2 3/4.4.9.2
[Pressurizer - Unit 1]
[Pressurizer Heatup and Cooldown Limits -
Unit 2]
Relocated No See Appendix A, page 9.
3/4.4.12 3/4.4.4 PORV Block Valves 3.4.11 Yes-4 3/4.4.13 3/4.4.9.3
[Power Operated Relief Valves - Unit 1]
[Overpressure Protection Systems - Unit 2]
3.4.12 Yes-2 3/4.4.14 NA Unit 1 Only - Reactor Coolant Pump -
Starting 3.4.6 3.4.7 Yes-2 3/4.4.15 3/4.4.10 Reactor Coolant System Vents Relocated No See Appendix A, page 10.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 9 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 Safety Injection Tanks 3.5.1 Yes-3 3/4.5.2 ECCS Subsystems - Operating 3.5.2 3.5.5 (U2)
Yes-3 3/4.5.3 ECCS Subsystems - Shutdown 3.5.3 Yes-3 3/4.5.4 Refueling Water Tank 3.5.4 Yes-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL 3/4.6.1.1
[Containment Vessel Integrity - Unit 1]
[Containment Integrity - Unit 2]
3.6.1 Yes-3 ISTS 3.6.1B 3/4.6.1.2 Containment Leakage 3.6.1 Yes-3 ISTS 3.6.1B Containment leakage is being retained as a Surveillance Requirement (SR 3.6.1.1) in ITS 3.6.1.
3/4.6.1.3 Containment Air Locks 3.6.2 Yes-3 3/4.6.1.4 Internal Pressure 3.6.4 Yes-2 ISTS 3.6.4B
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 10 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.6.1.5 Air Temperature 3.6.5 Yes-2 3/4.6.1.6 Containment Vessel Structural Integrity 3.6.1 Yes-3 ISTS 3.6.1B Containment vessel structural integrity is being retained as a Surveillance Requirement (SR 3.6.1.1) in ITS 3.6.1.
NA 3/4.6.1.7 Unit 2 only - Containment Ventilation System 3.6.3 Yes-3 ISTS 3.6.3 addresses 48 inch purge and exhaust valves.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 Containment Spray and Cooling Systems 3.6.6 Yes-3 ISTS 3.6.6A 3/4.6.2.2 NA Unit 1 only - Spray Additive System 3.6.10 Yes-3 ISTS 3.6.7 3/4.6.3 Containment Isolation Valves 3.6.3 Yes-3 3/4.6.5 Vacuum Relief Valves 3.6.8 Yes-3 ISTS 3.6.12 3/4.6.6 SECONDARY CONTAINMENT 3/4.6.6.1 Shield Building Ventilation System 3.6.9 Yes-3 For Unit 2: Combines ISTS 3.6.8 and ISTS 3.7.14 3/4.6.6.2 Shield Building Integrity 3.6.7 Yes-3 ISTS 3.6.11
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 11 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.6.6.3 Shield Building Structural Integrity 3.6.7 Yes-3 ISTS 3.6.11 Shield building structural integrity is being retained as a Surveillance Requirement; SR 3.6.7.3 in ITS 3.6.7.
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 Safety Valves 3.7.1 Yes-3 3/4.7.1.2 Auxiliary Feedwater System 3.7.5 Yes-3 3/4.7.1.3 Condensate Storage Tank 3.7.6 Yes-2, 3 3/4.7.1.4 Activity 3.7.1.16 Yes-2 ISTS 3.7.19 3/4.7.1.5 Main Steam Line Isolation Valves 3.7.2 Yes-3 NA 3/4.7.1.6 Main Feedwater Isolation Valves 3.7.3 Yes-3 Adding ISTS 3.7.3, Main Feedwater Isolation Valves (MFIVs), to Unit 1 ITS.
NA 3/4.7.1.7 Atmospheric Dump Valves 3.7.4 Yes-3 Adding ISTS 3.7.4, Atmospheric Dump Valves (ADVs), to Unit 1 ITS.
3/4.7.2.1 3/4.7.2 Steam Generator Pressure/Temperature Limitation Relocated No See Appendix A, page 11.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 12 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.7.3 Component Cooling Water System 3.7.7 Yes-3 3/4.7.4 Intake Cooling Water System 3.7.8 Yes-3 3/4.7.5 Ultimate Heat Sink 3.7.9 Yes-3 NA 3/4.7.6 Unit 2 only - Flood Protection Relocated No See Appendix A, page 12.
General requirements associated with hazard barriers, including flooding barriers, that affect Technical Specification systems, trains, subsystems, and components are included in ITS LCO 3.0.9, as modified consistent with NUREG-1432, Revision 5.
3/4.7.7
[Control Room Emergency Ventilation System - Unit 1]
[Control Room Emergency Air Cleanup System (CREACS) - Unit 2]
3.7.10 3.7.11 Yes-3 ISTS 3.7.11 ISTS 3.7.12 3/4.7.8 ECCS Area Ventilation System 3.7.12 Yes-3 ISTS 3.7.13 3/4.7.9 3/4.7.10 Sealed Source Contamination Relocated No See Appendix A, page 13.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 13 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.7.10 3/4.7.9 Snubbers Deleted No General requirements associated with snubbers that affect Technical Specification systems, trains, subsystems, and components are included in ITS LCO 3.0.8, as modified consistent with NUREG-1432, Revision 5.
See technical change discussions related to Unit 1 CTS 3/4.7.10 and Unit 2 CTS 3/4.7.9 in the Discussion of Changes for ITS Sections 3.0 and 3.7.
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 3/4.8.1.1 Operating 3.8.1 3.8.3 Yes-3 3/4.8.1.2 Shutdown 3.8.2 3.8.3 Yes-3 NA 3/4.8.2 D.C. SOURCES 3/4.8.2.3 3/4.8.2.1
[D.C. Distribution - Operating - Unit 1]
[Operating - Unit 2]
3.8.4 3.8.6 Yes-3 3/4.8.2.4 3/4.8.2.2
[D.C. Distribution - Shutdown - Unit 1]
[Shutdown - Unit 2]
3.8.5 3.8.6 Yes-3
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 14 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.8.2 3/4.8.3 ONSITE POWER DISTRIBUTION
[SYSTEMS - Unit 1]
3/4.8.2.1 3/4.8.3.1
[A.C. Distribution - Operating - Unit 1]
[Operating - Unit 2]
3.8.7 3.8.9 Yes-3 Adding ISTS 3.8.7, Inverters - Operating, to Unit 1 ITS 3/4.8.2.2 3/4.8.3.2
[A.C. Distribution - Shutdown - Unit 1]
[Shutdown - Unit 2]
3.8.8 3.8.10 Yes-3 Adding ISTS 3.8.7, Inverters - Shutdown, to Unit 1 ITS 3/4.8.2.3 3/4.8.3.1
[D.C. Distribution - Operating - Unit 1]
[Operating - Unit 2]
3.8.9 Yes-3 3/4.8.2.4 3/4.8.3.2
[D.C. Distribution - Shutdown - Unit 1]
[Shutdown - Unit 2]
3.8.10 Yes-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 Boron Concentration 3.9.1 Yes-2 3/4.9.2 Instrumentation 3.9.2 Yes-3 3/4.9.3 Decay Time Deleted No See CTS 3/4.9.3 technical change discussion in the Discussion of Changes for ITS Section 3.9.
3/4.9.4 Containment Penetrations 3.9.3 Yes-3
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 15 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION 3/4.9.8.1 High Water Level 3.9.4 Yes-2 3/4.9.8.2 Low Water Level 3.9.5 Yes-2 3/4.9.9 Containment Isolation System 3.9.3 Yes-3 3/4.9.10 Water Level - Reactor Vessel 3.9.6 Yes-2 3/4.9.11 Spent Fuel Storage Pool 3.7.13 3.7.14 Yes-2, 3 Yes-2 ISTS 3.7.16 ISTS 3.7.17 3/4.9.12 NA Unit 1 only - Fuel Pool Ventilation System -
Fuel Storage 3.7.17 Yes-3 ISTS 3.7.14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 Shutdown Margin Deleted No Compliance with an STE is optional; therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. See CTS 3/4.10.1 technical change discussion in the Discussion of Changes for ITS Section 3.1.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 16 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 3/4.10.2
[Moderator Temperature Coefficient - Unit 2]
Group Height, Insertion and Power Distribution Limits 3.1.8 Yes Compliance with an STE is optional; therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. This Specification has been retained to provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs.
NA 3/4.10.3 Unit 2 only - Reactor Coolant Loops Deleted.
No Compliance with an STE is optional; therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. See U2 CTS 3/4.10.3 technical change discussion in the Discussion of Changes for ITS Section 3.4.
3/4.10.5 3/4.10.4 Center CEA Misalignment Deleted No Compliance with an STE is optional; therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. See U1 CTS 3/4.10.5 and U2 CTS 3/4.10.4 technical change discussion in the Discussion of Changes for ITS Section 3.1.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 17 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes NA 3/4.10.5 Unit 2 only - CEA Insertion During ITC, MTC, and Power Coefficient Measurements 3.1.8 Yes Not explicitly included in NUREG-1432.
ISTS 3.1.8 limits power to 85% RTP.
Compliance with an STE is optional; therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. This Specification has been retained to provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs.
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.5 Explosive Gas Mixture 5.5 Yes Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W.
T. Russell to the industry ITS Chairpersons, dated October 25, 1993.
3/4.11.2.6 Gas Storage Tanks 5.5 Yes Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W.
T. Russell to the industry ITS Chairpersons, dated October 25, 1993.
SUMMARY
DISPOSITION MATRIX FOR ST. LUCIE PLANT (PSL) UNIT 1 AND UNIT 2 (1) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 18 of 18 U1 Current TS (CTS)
Number U2 CTS Number (if different)
CTS Specification Title New TS (ITS)
Reference Retained/
Criterion For Inclusion (1)
Notes 5.0 DESIGN FEATURES 3.7.15 4.0 Yes-2 Yes ISTS 3.7.18 Application of Technical Specification selection criteria is not appropriate.
However, specific portions of Design Features will be included in Technical Specifications as required by 10 CFR 50.36.
6.0 ADMINISTRATIVE CONTROLS 5.0 Yes Application of Technical Specification selection criteria is not appropriate.
However, specific portions of Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36.
APPENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 1 of 13 Unit 2 CTS Table 2.2-1, Reactor Protective System Instrumentation Trip Setpoint Limits, Functional Unit 10 Unit 2 CTS Tables 3.3-1 and 4.3-1, Reactor Protective System Instrumentation, Functional Unit 10 DISCUSSION:
The Reactor Protective System (RPS) initiates a reactor trip to protect against violating the core specified acceptable fuel design limits and breaching the reactor coolant pressure boundary (RCPB) during anticipated operational occurrences (AOOs). By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents. The protection and monitoring systems 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 Limiting Conditions for Operation on other reactor system parameters and equipment performance.
Technical Specifications are required by 10 CFR 50.36 to include LSSS for variables that have significant safety functions.
The NRC position on application of the screening criteria to RPS instrumentation is documented in a letter dated May 9, 1988 from T.E. Murley (NRC) to the C-E Owners Group (Reference 2). The Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function specified in CTS Table 2.2-1, 3.3-1 and 4.3-1 is provided as a plant specific reactor trip as an equipment protective feature required as a stipulation by NRC staff in justifying acceptance of the Reactor Coolant Water Component Cooling Water piping design. This reactor trip is not required for reactor protection and is not assumed in any accident or transient analysis. In addition, this automatic reactor trip feature on a loss of component cooling water (CCW) flow to the reactor coolant pumps (RCPs) is not required to protect against violating the core specified acceptable fuel design limits or breaching the RCPB during AOOs and is not considered an LSSS for a variable that has a significant safety function.
COMPARISON TO SELECTION CRITERIA:
- 1. The Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.
- 3. The Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function is not a structure, system, or component that is part of the primary success path or which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. The loss of CCW flow to the RCPs is addressed in the PSL PRA with respect to requiring operator action to recover the CCW flow or remove the RCPs from service. The Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function is not a significant risk contributor to core damage frequency and offsite releases in the PSL PRA. Therefore, the Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function does not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 2 of 13 CONCLUSION:
Since the selection criteria have not been satisfied for Loss of Component Cooling Water to Reactor Coolant Pumps - Low automatic reactor trip function, the requirements of this reactor trip instrument may be relocated to a licensee controlled document outside the Technical Specifications. Other reactor trip functions that are necessary to protect against violating the core specified acceptable fuel design limits or breaching the RCPB during AOOs are retained in the Technical Specifications (e.g., Reactor Coolant Flow - Low reactor trip and manual reactor trip functions).
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 3 of 13 Current Technical Specification (CTS) 3/4.1.1.3, Boron Dilution DISCUSSION:
CTS 3/4.1.1.3 establishes a minimum flow rate to the reactor pressure vessel (RPV) of at least 3000 gpm whenever a reduction in Reactor Coolant System (RCS) boron concentration is being made.
This minimum flowrate provides adequate mixing, prevents stratification, and ensures that reactivity changes will be gradual during boron concentration changes in the RCS.
A minimum boron dilution flow rate to the RPV is not assumed as an initial condition of a design basis accident (DBA) or transient analysis nor is it an assumed value to mitigate a DBA or transient (e.g., a boron dilution event).
COMPARISON TO SELECTION CRITERIA:
- 1. The minimum boron dilution flow rate to the RPV is not an instrumentation system. Therefore, this Specification does not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The minimum boron dilution flow rate to the RPV is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This Specification specifies a minimum flowrate value that provides adequate mixing, prevents stratification, and ensures that reactivity changes in the core are gradual during boron concentration changes in the RCS. These limits, however, do not reflect initial condition assumptions in a DBA or transient.
- 3. The minimum boron dilution flow rate to the RPV is a parameter and not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. The minimum boron dilution flow rate to the RPV is a parameter requirement and is not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied, the Boron Dilution Specification may be relocated to a licensee controlled document outside the Technical Specifications. Control element assembly insertion limits during power operation, shutdown margin requirements during shutdown, and boration concentration requirements during refueling are retained in separate Technical Specifications and ensure adequate excess negative core reactivity is available in the event of an inadvertent boron dilution event. Additionally, RCS circulation requirements to provide mixing and prevent stratification are retained in separate Technical Specifications (i.e., shutdown cooling specifications).
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 4 of 13 CTS 3/4.1.2.1, Flow Paths - Shutdown; CTS 3/4.1.2.2, Flow Paths - Operating; CTS 3/4.1.2.3, Charging Pumps - Shutdown; CTS 3/4.1.2.4, Charging Pumps - Operating; CTS 3/4.1.2.5, Boric Acid Pumps - Shutdown; CTS 3/4.1.2.6, Boric Acid Pumps - Operating; CTS 3/4.1.2.7, Borated Water Sources - Shutdown; CTS 3/4.1.2.8, Borated Water Sources - Operating DISCUSSION:
The components associated with the boration system Technical Specifications provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin during normal operations. To accomplish this functional requirement, the current boration system technical specifications require a source of borated water, one or more flow paths to inject borated water into the RCS, and either a charging pump or high pressure safety injection pump to provide the necessary charging head.
The RCS boration management system functions to control boron concentration and maintain shutdown margin are not assumed to be OPERABLE to mitigate the consequences of a DBA or transient. In the case of a malfunction of a component in the boration systems which causes a boron dilution event, response by the operator is to close the appropriate valves in the reactor makeup system. The plant response to a boron dilution event also includes control rod assembly movement and reactor trip features to ensure shutdown margin is maintained. The boration capabilities of the boration systems are not assumed to mitigate the boron dilution event.
COMPARISON TO SELECTION CRITERIA:
- 1. The boration systems do not constitute instrumentation systems that are used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The boration systems are not process variables, design features, or operating restrictions that represent an initial condition of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.
- 3. The RCS boration management system functions to control boron concentration and maintain shutdown margin do not represent a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. The PSL at-power PRA has shown that the RCS boration management system functions are not significant risk contributors to core damage frequency and offsite releases. PSL does not have shutdown PRA model. However, operational experience has shown that the boration management system is not a constraint of prime importance in the mitigation of any accident or transient that results in challenging public health and safety. Therefore, the RCS boration management system functions to control boron concentration and maintain shutdown margin do not represent structures, systems, or components which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 5 of 13 CONCLUSION:
Since the selection criteria have not been satisfied for the RCS boration management system functions to control boron concentration and maintain shutdown margin, the Boration System Specifications may be relocated to a licensee controlled document outside the Technical Specifications. RCS boration structures, systems, or components credited as the primary success path which function or actuate to mitigate a DBA or transient are retained in separate Technical Specifications (e.g., Emergency Core Cooling System). Additionally, shutdown margin requirements during shutdown, and boration concentration requirements during refueling are retained in separate Technical Specifications to ensure adequate excess negative core reactivity is available in the event of an inadvertent boron dilution event.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 6 of 13 Unit 2 CTS 3/4.1.3.3, Position Indication Channels - Shutdown DISCUSSION:
When the reactor is shutdown, one CEA position indication channel for each withdrawn CEA must be OPERABLE when the reactor trip breakers are closed and one or more CEAs are withdrawn. The purpose of this requirement is to provide the control room operator with indication of the position of a CEA when the CEA is not fully inserted and perform any related operations that are required. The CEA position indication channels do not provide any automatic function and no operator action assumed in accident or transient analyses (e.g., uncontrolled CEA withdrawal event) is initiated based on CEA position. Shutdown margin requirements ensure adequate excess negative reactivity is available to maintain the reactor subcritical when a CEA with the highest reactivity worth is withdrawn while the reactor is shutdown.
COMPARISON TO SELECTION CRITERIA:
- 1. The CEA position monitoring system when the reactor is shutdown is not an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The CEA position monitoring system when the reactor is shutdown is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.
- 3. The CEA position monitoring system when the reactor is shutdown is not a structure, system, or component that is part of the primary success path or which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. The CEA position monitoring system when the reactor is shutdown and the reactor trip breakers are closed is not addressed in the PSL PRA and does not represent a system which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied for the CEA position monitoring system when the reactor trip breakers are closed and one or more CEAs are withdrawn, the requirements of the CEA Position Indication - Shutdown Specification may be relocated to a licensee controlled document outside the Unit 2 Technical Specifications. Shutdown Margin requirements during shutdown are retained in separate Technical Specifications and ensure adequate excess negative core reactivity is available in the event of a CEA system malfunction during shutdown (e.g., uncontrolled CEA withdrawal event from a subcritical condition).
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 7 of 13 Unit 1 CTS Table 3.3-11, Accident Monitoring Instrumentation, Instruments 2 through 6, and 8 Unit 2 CTS Table 3.3-10, Accident Monitoring Instrumentation, Instruments 6, 7, and 9 through 15 DISCUSSION:
Accident monitoring instrumentation ensures sufficient information is available following an accident to allow an operator to verify the response of automatic safety systems, and to take preplanned manual actions to accomplish a safe shutdown of the plant. The NRC position on application of the screening criteria to post-accident monitoring instrumentation is documented in a letter dated May 9, 1988 from T.E. Murley (NRC) to the C-E Owners Group (Reference 2). Regulatory Guide (RG) 1.97, Type A variables provide primary information; i.e., information that is essential for the direct accomplishment of the specified manual actions (including long term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for DBAs or transients. Additionally, it could not be confirmed that RG 1.97, non-Type A Category 1 variables are not of prime importance in limiting risk. Therefore, the NRC position is that the post-accident monitoring instrumentation list should contain Type A instruments and non-Type A Category 1 instruments specified in the plant's safety evaluation report (SER) on RG 1.97. Accordingly, this position has been applied to the PSL Units 1 and 2 RG 1.97 instruments.
A review of the PSL Units 1 and 2 UFSAR and the NRC Regulatory Guide 1.97 Safety Evaluation for PSL Units 1 and 2 shows that the following Unit 1 CTS Table 3.3-11 and Unit 2 CTS Table 3.3.10 Instruments do not meet Category 1 or Type A requirements:
Unit 1 CTS Table 3.3-11 Instrument 2 Auxiliary Feedwater Flow Rate Instrument 3 RCS Subcooling Margin Monitor Instrument 4 PORV Position Indicator - Acoustic Flow Monitor Instrument 5 PORV Block Valve Position Indicator Instrument 6 Safety Valve Position Indicator Instrument 8 Containment Sump Level Unit 2 CTS Table 3.3-10 Instrument 6 Steam Generator Pressure Instrument 7 Steam Generator Water Level - Narrow Range Instrument 9 Refueling Water Tank Water Level Instrument 10 Auxiliary Feedwater Flow Rate Instrument 11 RCS Subcooling Margin Monitor Instrument 12 PORV Position/Flow Indicator Instrument 13 PORV Block Valve Position Indicator Instrument 14 Safety Valve Position/Flow Indicator Instrument 15 Containment Sump Water Level (Narrow Range)
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 8 of 13 COMPARISON TO SELECTION CRITERIA:
- 1. These instruments are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The monitored parameters are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.
- 3. These instruments are not structures, systems, or components that are part of the primary success path or which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. These post-accident monitoring instrument functions are not addressed in the PSL PRA and do not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied for instruments which do not meet RG 1.97 Type A or non-Type A, Category 1 variable requirements, the requirements of these instruments may be relocated to a licensee controlled document outside the Technical Specifications. Post-accident monitoring instruments that monitor RG 1.97 Type A variables and non-Type A, Category 1 variables are retained in the Technical Specifications.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 9 of 13 Unit 1 CTS 3/4.4.9.2, Pressurizer Unit 2 CTS 3/4.4.9.2, Pressurizer Heatup and Cooldown Limits DISCUSSION:
The limitation on the pressurizer pressure and temperature (P/T) ensures that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. As such, the Technical Specification places limits on variables consistent with structural analysis results. These limits do not represent an initial condition assumption of a DBA or transient. Although the limits represent operating restrictions, and Criterion 2 includes operating restrictions, the Criterion 2 discussion of the Final Policy Statement specified that only those operating restrictions required to preclude unanalyzed accidents and transients be included in Technical Specifications.
COMPARISON TO SELECTION CRITERIA:
- 1. Pressurizer P/T limits are ASME Code parameter limits and are not instrumentation used for, or capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. Pressurizer P/T limits are ASME Code parameter limits and do not represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Pressurizer P/T limits are ASME Code parameter limits and do not represent a structure, system or component that is part of a primary success path in the mitigation of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. Pressurizer P/T limits are ASME Code parameter limits and do not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied, the Unit 1 Pressurizer Specification, which specifies pressurizer heatup and cooldown limits, and the Unit 2 Pressurizer Heatup and Cooldown Limits Specification may be relocated to a licensee controlled document outside the Technical Specifications.
Pressure and temperature limits associated with the limiting RCS pressure boundary components are retained in separate Technical Specifications. ASME code requirements associated with the limits on pressurizer heatup and cooldown will continue to be controlled pursuant 10 CFR 50.55a.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 10 of 13 Unit 1 CTS 3/4.4.15, Reactor Coolant System Vents Unit 2 CTS 3/4.4.10, Reactor Coolant System Vents DISCUSSION:
Reactor Coolant System (RCS) vents are provided to exhaust non-condensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The function, capabilities, and testing requirements of the RCS vent system are consistent with the requirements of Item II.b.1 of NUREG-0737, Clarification of TMI [Three Mile island] Action Plan Requirements, November 1980.
The pressurizer safety valves provide primary relief capability to exhaust non-condensible gases and steam from the RCS during a DBA or transient at power. The pressurizer power operated relief valves (PORVs) and other vent paths (e.g., pressurizer manway cover removed) provide vent capability during RCS low temperature and pressure conditions.
COMPARISON TO SCREENING CRITERIA:
- 1. The RCS vents do not function as installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The RCS vents are not process variables, design features, or operating restrictions that represents an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The RCS vents are not a structures, systems or components that are part of the primary success path and which 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.
- 4. The RCS vent function to preclude the inhibition of natural circulation core cooling is not a significant risk contributor to core damage frequency and offsite releases and this TMI function does not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the screening criteria have not been satisfied for the RCS vent function that precludes the inhibition of natural circulation core cooling, the Reactor Coolant System Vents Specification may be relocated to a licensee controlled document outside Technical Specifications. Pressure relief requirements during power operation and during low temperature and pressure conditions are retained in separate Technical Specifications and ensure relief capability is provided to exhaust non-condensible gases and steam from the primary system. In addition, this NUREG-0737 requirement is required pursuant to 10 CFR 50.34(f)(2)(vi). Compliance with applicable portions of 10 CFR 50.34(f)(2) is required by the operating licenses of PSL Units 1 and 2.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 11 of 13 CTS 3/4.7.2, Steam Generator Pressure/Temperature Limitation DISCUSSION:
The limitation on steam generator pressure and temperature (P/T) ensures that pressure-induced stresses on the steam generators do not exceed the maximum allowable fracture toughness limits.
These pressure and temperature limits are based on maintaining a steam generator reference temperature for Nil Ductility Transition (RTNDT) sufficient to prevent brittle fracture. As such, the Technical Specification places limits on variables consistent with structural analysis results. These limits do not represent an initial condition assumption of a DBA or transient. Although the limits represent operating restrictions, and Criterion 2 includes operating restrictions, the Criterion 2 discussion of the Final Policy Statement specified that only those operating restrictions required to preclude unanalyzed accidents and transients be included in Technical Specifications.
COMPARISON TO SELECTION CRITERIA:
- 1. Steam generator P/T limits are ASME Code parameter limits and are not instrumentation used for, or capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. Steam generator P/T limits are ASME Code parameter limits and do not represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Steam generator P/T limits are ASME Code parameter limits and not a structure, system, or component that is part of a primary success path in the mitigation of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. Steam generator P/T limits are ASME Code parameter limits and are not structures, systems, or components which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied, the Steam Generator Pressure/Temperature Limitation Specification may be relocated to a licensee controlled document outside the Technical Specifications. Pressure and temperature limits associated with the limiting RCS pressure boundary components are retained in separate Technical Specifications. ASME code requirements associated with the limits on steam generator heatup and cooldown will continue to be controlled pursuant 10 CFR 50.55a
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 12 of 13 Unit 2 CTS 3/4.7.6, Flood Protection DISCUSSION:
External flooding during hurricane conditions is not designated as a PSL DBA or transient event. In addition, external flooding due to a hurricane is not postulated to occur during any DBA or transient, thus water level (as it pertains to wave run-up effects during a hurricane) is not credited in any safety analysis. The Flood Protection Technical Specification requirements ensure that facility protective actions will be taken in the event of flood conditions whenever a hurricane warning is issued. The installation of the stoplogs ensures adequate protection for wave run-up effects where no permanent adjacent structures exist and provides protection to safety-related equipment.
COMPARISON TO SCREENING CRITERIA:
- 1. Flood protection requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. Flood protection 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.
- 3. Flood protection 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.
- 4. The specific requirements specified in the Flood Protection Specification related to a hurricane are found to be non-significant risk contributors to core damage frequency and offsite releases and do not represent structures, systems, or components which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the screening criteria have not been satisfied, the Flood Protection Specification may be relocated to a licensee controlled document outside the Unit 2 Technical Specifications. A general requirement associated with hazard barriers (ITS LCO 3.0.9), which includes external flooding barriers, is added to ITS Section 3.0. A hazard barrier that cannot perform its related support function will be evaluated and managed under the Maintenance Rule plant configuration control requirement, 10 CFR 50.65(a)(4), and associated industry guidance, NUMARC 93-01, Revision 4A.
APPLICATION OF SELECTION CRITERIA TO THE ST LUCIE PLANT UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS Appendix A - Justification for Specification Relocation Page 13 of 13 Unit 1 CTS 3/4.7.9, Sealed Source Contamination Unit 2 CTS 3/4.7.10, Sealed Source Contamination DISCUSSION:
The limitations on sealed source contamination are intended to ensure that the total body and individual organ irradiation doses do not exceed allowable limits in the event of ingestion or inhalation. This is done by imposing a maximum limitation of < 0.005 microcuries of removable contamination on each sealed source. This requirement and the associated surveillance requirements bear no relation to the conditions or limitations that are necessary to ensure safe reactor operation.
COMPARISON TO SELECTION CRITERIA:
- 1. The sealed source contamination limitation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The sealed source contamination limitation is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The sealed source contamination limitation is not part of a primary success path in the mitigation of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 4. The sealed source contamination limitation is not discussed in the PSL PRA and is not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
CONCLUSION:
Since the selection criteria have not been satisfied, the Sealed Source Contamination Specification may be relocated to a licensee controlled document outside the Technical Specifications. Requirements associated with the sealed sources are governed by 10 CFR Part 70. Compliance with applicable portions of 10 CFR Part 70 is required by the operating licenses of PSL Units 1 and 2.