ML24142A430
| ML24142A430 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/20/2024 |
| From: | Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24142A428 | List:
|
| References | |
| LR-N24-0029, LAR H24-02 | |
| Download: ML24142A430 (1) | |
Text
LR-N 24-0029 LAR H24-0 2
ENCLOSURE 2
HOPE CREEK GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS SUBMITTAL
VOLUMES 1 THROUGH 17
(3379 TOTAL PAGES, INCLUDING COVER SHEETS)
ENCLOSURE 2
VOLUME 1
HOPE CREEK GENERATING STATION
IMPROVED TECHNICAL SPECIFICATIONS CONVERSION
APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Revision 0
APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Revision 0
APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
CONTENTS Page
- 1. INTRODUCTION......................................................................................................... 1
- 2. SELECTION CRITERIA............................................................................................... 2
- 3. PRA INSIGHTS........................................................................................................... 4
- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA........................................ 7
- 5. REFERENCES........................................................................................................... 8
ATTACHMENT
- 1.
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
APPENDIX
A. JUSTIFICATION FOR SPECIFICATION RELOCATION
i APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
- 1. INTRODUCTION
The purpose of this document is to confirm the results of the Boiling Water Reactor Owners Group (BWROG) application of the Technical Specification selection criteria on a plant specific basis for the Hope Creek Generating Station (Hope Creek). PSEG Nuclear LLC (PSEG ) has reviewed the application and confirmed the applicability of the selection criteria to e ach of the Technical Specifications utilized in report NEDO -31466, "Technical Specification Screening Criteria Application and Risk Assessment," including Supplement 1 (References 1 and 2), NRC Staff Review of N uclear 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 3), and as revised in NUREG -1433,
Revision 5.0 "Standard Technical Specifications, General Electric Plants, BWR/4" (Reference 4 )
and applied the criteria to each of the current Hope Creek Technical Specifications.
Additionally, in accordance with the NRC guidance, this confirmation of the application of selection criteria includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in Reference 2, as applicable to the Hope Creek.
Page 1 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
- 2. SELECTION CRITERIA
PSEG 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 6) to develop the results contained in the attached matrix. PRA insights as used in the BWROG submittal were utilized, confirmed by PSEG, and are discussed in the next section of this report. The selection criteria and discussion provided in Reference 6 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 Specific ations 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 ev ents, analyzed in the Final Safety Analysis Report (FSAR), for which a structure, system, or component must meet specified functional goals. These analyses are contained in Chapters 6 and 15 of the UFSAR (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
Page 2 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
power operation. As long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure/low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.
Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a 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 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 plants 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 D BAs and T ransients 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 primar y 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 licensees retain in their Technical Specifications LCOs, action statements and Surveillance Requirements for the following systems, which are 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 Policies, be retained or included in Technical Specifications.
Page 3 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION 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 submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to mak e better use of risk and reliability information for defining future generic Technical Specification requirements.
- 3. PRA INSIGHTS
Introduction and Objectives
The Final Policy Statement (Reference 6) 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 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. I n addition, in some cases risk was judged independent of any specific PR A 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 documented in Ref erences 1 and 2, except for Specifications discussed in Appendix A, "Justification for Specification Relocation, " and has been confirmed by PSEG for those Specifications to be relocated.
Assumptions and Approach
To summarize, the approach used in References 1 and 2 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 c alculation 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. List the function(s) affected by removal of the LCO item.
- b. Determine the effect of loss of the LCO item on the function(s).
Page 4 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
- c. Identify compensating provisions, redundancy, and backups related to the loss of the LCO item.
- d. 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.
- e. 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.
- f. 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
Frequency High Medium Low High S S NS
Medium S S NS Low NS NS NS
S = Potential Significant Risk Contributor NS = Risk Non-Significant
- g. 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 5 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
TABLE 3-1 BWR PRAs USED IN NED O-31466 (and Supplement 1)
RISK ASSESSMENT
- BWR/6 Standard Plant, GESSAR II, 238 Nuclear Island, BWR/6 Standard Plant Probabilistic Risk Assessment, Docket No. STN 50-447, March 1982.
- La Salle County Station, NED O-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, SAl-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-1659, "Reactor Safety Study Methodology Applications Program:
Grand Gulf #1 BWR Power Plant," October 1981.
- NEDC-30936P, "BWR Owners' Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 2," June 1987.
Page 6 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA
The selection criteria from Section 2 were applied to the Hope Creek Technical Specifications.
Attachment 1 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 selection criteria are provided in Appendix A, except as noted in the Summary Disposition Matrix.
Page 7 of 8 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
- 5. REFERENCES
- 1. NEDO-31466, "Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
- 2. NEDO-31466, Supplement 1, "Technical Specification Screening Criteria Application and Risk Assessment," February 1990.
- 3. 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.
(ADAMS Accession No. ML11264A057)
- 4. NUREG-1433, "Standard Technical Specifications, General Electric BWR/4 Plants" Revision 5.0, September 2021.
- 5. 10 CFR 50.36, Technical specifications, (c)(2)(ii) selection criteria, July 19, 1995 (Federal Register Notice 60 FR 36953).
- 6. Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, July 22, 1993 (Federal Register Notice 58 FR 39132).
Page 8 of 8 ATTACHMENT 1
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
1.0 DEFINITIONS 1.1 Yes/NA This section provides definitions for terms used throughout the remainder of Technical Specifications.
Definitions are provided to explain the meaning of certain terms. As such, direct application of the Technical Specification selection criteria is not appropriate. However, only those definitions for defined terms that remain as a result of application of the selection criteria will remain as definitions in this section of Technical Specifications.
2.1 SAFETY LIMITS
2.1.1 Thermal Power, Low Pressure or 2.1.1.1 Yes/NA Application of Technical Specification selection criteria Low Flow is not appropriate. However. Safety Limits and Limiting Safety System Settings (as part of Reactor Protection 2.1.2 Thermal Power, High Pressure and 2.1.1.2 System Instrumentation) will be included in Technical High Flow Specifications as required by 10 CFR 50.36.
2.1.3 Reactor Coolant System Pressure 2.1.2
2.1.4 Reactor Vessel Water Level 2.1.1.3
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
2.2 LIMITING SAFETY SYSTEM SETTINGS
2.2.1 Reactor Protection System 3.3.1.1 Yes/NA The Application of Technical Specification selection Instrumentation Setpoints criteria is not appropriate. However, the Reactor Protection System (RPS) limited safety system settings (LSSS) have been included as part of the RPS instrumentation Specification, which has been retained because the Instrumentation 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.
3.0 APPLICABILITY LIMITING CONDITION FOR OPERATION
3.0.1 Operational Conditions LCO 3.0.1 Yes/NA This Specification provides generic guidance applicable to one or more Specifications. The 3.0.2 Noncompliance LCO 3.0.2 information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance 3.0.3 Generic Actions LCO 3.0.3 Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 3.0.4 Entry into Operational Conditions LCO 3.0.4 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG - 1433, Revision 5.
3.0.5 Equipment Removal from Service LCO 3.0.5
3.0.8 Inoperability of Snubbers LCO 3.0.8
3.0.9 Barriers LCO 3.0.9
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS
4.0.1 Operational Conditions SR 3.0.1 Yes/NA These Specifications provide generic guidance applicable to one or more Specifications. The 4.0.2 Time of Performance SR 3.0.2 information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance 4.0.3 Noncompliance SR 3.0.3 Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate. However, the general requirements of 4.0.4 Entry into Operational Conditions SR 3.0.4 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1433, Revision 5.
3/4.1 REACTIVITY CONTROL SYSTEMS
3/4.1.1 Shutdown Margin 3.1.1 Yes -2 Not a measured process variable but is important parameter used to confirm the acceptability of the accident analysis. In addition, LCO is retained as directed by the NRC.
3/4.1.2 Reactivity Anomalies 3.1.2 Yes -2 Confirms assumptions made in the reload safety analysis.
3/4.1.3 CONTROL RODS
3/4.1.3.1 Control Rod Operability 3.1.3 Yes-3 Control rods are part of the primary success path in 3.1.8 mitigating the consequences of design basis accidents (DBAs) and transients. The scram discharge volume vent and drain valves contribute to the operability of the control rod scram function.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.1.3.2 Control Rod Maximum Scram 3.1.3 Yes-3 Control rods are part of the primary success path in Insertion Times mitigating the consequences of DBAs and transients.
3/4.1.3.3 Control Rod Scram Insertion Times 3.1.4 Yes-3 Control rods are part of the primary success path in mitigating the consequences of DBAs and transients.
3/4.1.3.5 Control Rod Scram Accumulators 3.1.5 Yes-3 Control rods are part of the primary success path in 3.9.5 mitigating the consequences of DBAs and transients.
3/4.1.3.6 Control Rod Drive Coupling 3.1.3 Yes-3 Control rods are part of the primary success path in mitigating the consequences of DBAs and transients.
3/4.1.3.7 Control Rod Position Indication 3.1.3 Yes-3 Control rod positions are part of the primary success 3.9.4 path in mitigating the consequences of DBAs and transients.
3/4.1.3.8 Control Rod Drive Housing Support Deleted No See technical change discussion in the CTS 3/4.1.3.8 Discussion of Changes for the Control Rod Drive Housing Support.
3/4.1.4 Control Rod Program Controls
3/4.1.4.1 Rod Worth Minimizer 3.3.2.1 Yes -3 Prevents withdrawal of out -of -sequence control rods that might set - up high rod worth conditions beyond Control Rod Drive Assembly ( CRDA ) assumptions.
3/4.1.4.3 Rod Block Monitor 3.3.2.1 Yes -3 Prevents continuous withdrawal of a high worth control rod that would challenge the MCPR Safety Limit and 1 percent cladding strain fuel design limit.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.1.5 Standby Liquid Control System 3.1.7 Yes-4 Retained in accordance with the NRC Final Policy Statement on Technical Specification improvements due to risk significance.
3/4.2 POWER DISTRIBUTION LIMITS
3/4.2.1 Average Planar Linear Heat 3.2.1 Yes-2 Peak cladding temperature following a LOCA is Generation Rate primarily dependent on initial APLHGR. As such, it is an initial condition of a DBA analysis.
3/4.2.3 Minimum Critical Power Ratio 3.2.2 Yes-2 Utilized as an initial condition of the design basis transients. Transient analys es are performed to establish the largest reduction in Critical Power Ratio.
This value is added to the fuel cladding integrity safety limit to determine the MCPR value.
3/4.2.4 Linear Heat Generation Rate 3.2.3 Yes-2 LHGR is calculated to avoid exceeding plastic strain limits on fuel rods. As such, it is an initial condition to Design Basis Transient Analyses.
3/4.3 INSTRUMENTATION
3/4.3.1 Reactor Protection System 3.3.1.1 Yes-3 Actuates to mitigate consequences of a DBA and/or Instrumentation transient, or because it provides an anticipatory scram to ensure the scram discharge volume and thus RPS remains operable, or it is retained as directed by the NRC as it is part of the RPS.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.3.2 Isolation Actuation Instrumentation 3.3.6.1 Yes-3 Actuates to mitigate the consequences of a DBA LOCA Table 3.3.2-1 3.3.6.2 or actuates to mitigate the consequences of a DBA Functions except LOCA release to the environment and a fuel handling Function 3.b accident, or actuates to isolate potential leakage paths to secondary containment consistent with safety analysis assumptions, or is retained due to risk significance, or is retained as directed by the NRC as it is part of the isolation system.
3/4.3.3 Main Steam Line Isolation on Main Relocated No See Appendix A, Page 1.
Table 3.3.3-1 Steam Line Radiation - High High Function 3.b
3/4.3.3 Emergency Core Cooling System 3.3.5.1 Yes-3 Actuates to mitigate the consequences of a DBA Table 3.3.3-1 Actuation Instrumentation LOCA, or is being retained due to risk significance, or Functions except functions to mitigate the consequences of a small break Functions 4.h and LOCA, or retained as required by the NRC as it is part 5 of the ECCS actuation system.
3/4.3.3 ADS Manual Inhibit Switch Relocated No See Appendix A, Page 3.
Table 3.3.3-1 Function 4.h
3/4.3.3 Loss of Power Instrumentation 3.3.8.1 Yes-3 Loss of power instrumentation actuates to assure Table 3.3.3-1 power availability to the ECCS and other safety-related Function 5 systems in the event of a loss of offsite power.
Mitigation of DBAs relies on the availability of the ECCS and other safety-related systems.
3/4.3.4.1 ATWS Recirculation Pump Trip 3.3.4.2 Yes-4 Retained in accordance with the NRC Final Policy System Instrumentation Statement on Technical Specification Improvements due to risk significance.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.3.4.2 End-Of-Cycle Recirculation Pump 3.3.4.1 Yes-3 EOC-RPT aids the reactor scram in protecting fuel Trip System Instrumentation cladding integrity by ensuring the fuel cladding integrity safety limit is not exceeded during a load rejection or turbine trip transient.
3/4.3.5 Reactor Core Isolation Cooling 3.3.5.3 Yes -3 RCIC is credited in the loss of feedwater transient System Actuation Instrumentation analysis to preclude reliance on the low pressure ECCS. Also, r etained in accordance with the NRC Final Policy Statement on Technical Specification Improvements due to risk significance or is retained as required by the NRC as it is part of the RCIC actuation system.
3/4.3.6 Control Rod Block Instrumentation 3.3.2.1 Yes -3 Prevents continuous withdrawal of a high worth control Table 3.3. 6 - 1 ( Table 3.3. 6 - 1 Function 1, Rod Block rod that would challenge the MCPR Safety Limit and Functions except Monitor and Function 7, Reactor one percent cladding plastic strain fuel design limit.
Functions 2, 3, 4, Mode Switch Shutdown Position )
and 5
3/4.3.6 APRM Relocated No See Appendix A, Page 4.
Table 3.3. 6 - 1 Function 2
3/4.3.6 Source Range Monitors Relocated No See Appendix A, Page 5.
Table 3.3. 6 - 1 Function 3
3/4.3.6 Intermediate Range Monitors Relocated No See Appendix A, Page 6.
Table 3.3. 6 - 1 Function 4
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.3.6 Scram Discharge Volume Relocated No See Appendix A, Page 7.
Table 3.3.6-1 Function 5
3/4.3.7 Monitoring Instrumentation 3.3.7
3/4.3.7.1 Radiation Monitoring Instrumentation 3.3.7.1 Yes-3 Actuates to maintain control room habitability so that Table 3.3.7.1-1 Control Room Ventilation Radiation operation can conti nue from the control room following Function 1 Monitoring DBAs.
3/4.3.7.1 2. Area Monitors Relocated No See Appendix A, Pages 8 and 9.
Table 3.3.7.1-1 a. Criticality Monitors Function 2 1) New Fuel Storage Vault
- 2) Spent Fuel Storage Pool
- b. Control Room Direct Radiation Monitor
3/4.3.7.1 Reactor Auxiliaries Cooling Relocated No See Appendix A, Page 10.
Table 3.3.7.1-1 Radiation Monitor Function 3
3/4.3.7.1 Safety Auxiliaries Cooling Relocated No See Appendix A, Page 11.
Table 3.3.7.1-1 Radiation Monitor Function 4
3/4.3.7.1 Offgas Pre-treatment Relocated No See Appendix A, Page 12.
Table 3.3.7.1-1 Radiation Monitor Function 5
3/4.3.7.4 Remote Shutdown System 3.3.3.2 Yes-4 Retained as directed by the NRC as it is a significant contributor to risk reduction.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.3.7.5 Accident Monitoring Instrumentation 3.3.3.1 Yes-3, 4 Regulatory Guide 1.97 Type A and Category 1 Table 3.3.7.5-1 (Type A and Category 1 Instruments) variables retained.
Functions except Functions 7, 11, 12, 13
3/4.3.7.5 Drywell Air Temperature Relocated No See Appendix A, Page 13.
Table 3.3.7.5-1 Function 7
3/4.3.7.5 North Plant Vent Radiation Monitor Table 3.3.7.5-1 Function 11
3/4.3.7.5 South Plant Vent Radiation Monitor Table 3.3.7.5-1 Function 12
3/4.3.7.5 FRVS Vent Radiation Monitor Table 3.3.7.5-1 Function 13
3.3.7.6 Source Range Monitors 3.3.1.2 Yes Does not satisfy the selection criteria, however, is being retained because the NRC considers it necessary for flux monitoring during shutdown, startup, and refueling operations.
3/4.3.9 Feedwater/Main Turbine Trip System 3.3.2.2 Yes -3 Actuates to limit feedwater addition to the reactor Actuation Instrumentation vessel on feedwater controller failure consistent with safety analysis assumptions. Limits neutron flux peak and thermal transient to avoid fuel damage.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.3.10 Mechanical Vacuum Pump Trip 3.3.7.2 Yes-3 New Specification 3.3.7.2, Mechanical Vacuum Pump Instrumentation Trip Instrumentation. There is no NUREG 1433 ISTS for Mechanical Vacuum Pump Trip Instrumentation.
Credited to mitigate the consequences of control rod drop accident (CRDA).
3/4.3.12 RPV Water Inventory Control 3.3.5.2 Yes-4 RPV water inventory control is required to protect the Instrumentation reactor vessel water level Safety Limit and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur.
3/4.4 REACTOR COOLANT SYSTEM
3/4.4.1 Recirculation System
3/4.4.1.1 Recirculation Loops 3.4.1 Yes-2 Recirculation loop flow is an initial condition in the 3.4.10 safety analysis.
3/4.4.1.2 Jet Pumps 3.4.2 Yes-2 Jet pump operability is assumed in the LOCA analysis to assure adequate core reflood capability.
3/4.4.1.3 Recirculation Loop Flow 3.4.1 Yes-2 Recirculation loop flow mismatch, within limits, is an initial condition in the safety analysis.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.4.1.4 Idle Recirculation Loop Startup 3.4.10 Yes-2 Temperature differential between the reactor coolant in the reactor vessel and the idle loop is an initial condition in the transient analysis. Idle loop startup with temperatures outside the limit could result in a reactivity transient and potential violation of the Safety Limit MCPR.
3/4.4.2 Safety/Relief Valves
3/4.4.2.1 Safety/Relief Valves 3.4.3 Yes-3 A minimum number of S/RVs is assumed in the safety analyses to mitigate overpressure events.
3/4.4.2.2 Safety/Relief Valves Low-Low Set 3.6.1.6 Yes-3 A minimum number of S/RVs is assumed in the Function 3.3.6.3 containment loading safety analysis.
3/4.4.3 Reactor Coolant System Leakage
3/4.4.3.1 Leakage Detection Systems 3.4.6 Yes-1 Leak detection is used to indicate a significant abnormal condition of the reactor coolant system pressure boundary.
3/4.4.3.2 Operational Leakage 3.4.4 Yes-2 Leakage beyond limits would indicate an abnormal 3.4.5 condition of the reactor coolant system pressure boundary. Operation outside of this condition is unanalyzed and is indicative of reactor coolant system pressure boundary failure.
3/4.4.5 Specific Activity 3.4.7 Yes-2 Specific activity provides an indication of the onset of significant fuel cladding failure and is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.4.6 Pressure/Temperature Limits
3/4.4.6.1 Reactor Coolant System 3.4.10 Yes-2 Establishes initial conditions to operation 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 system pressure boundary integrity.
3/4.4.6.2 Reactor Steam Dome 3.4.11 Yes-2 Reactor Steam Dome Pressure is an initial condition of the vessel overpressure protection analysis.
3/4.4.7 Main Steam Line Isolation Valves 3.6.1.3 Yes-3 Main steam line isolation within specified time limits ensures the release to the environment is consistent with the assumptions in the LOCA analysis.
3/4.4.9 Residual Heat Removal
3/4.4.9.1 Hot Shutdown 3.4.8 Yes-4 Retained in accordance with the NRC Final Policy on Technical Specification Improvements due to risk.
3/4.4.9.2 Cold Shutdown 3.4.9 Yes-4 Retained in accordance with the NRC Final Policy on Technical Specification Improvements due to risk.
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL
3/4.5.1 ECCS - Operating 3.5.1 Yes -3 Functions to mitigate the consequences of a DBA.
3/4.5.2 RPV Water Inventory Control 3.5.2 Yes -4 Functions to mitigate the consequences of a vessel draindown event.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.5.3 Suppression Chamber 3.6.2.2 Yes-2,3 Suppression pool water level is an initial condition in the DBA LOCA analysis and mitigates the consequences of a DBA.
3/4.6 CONTAINMENT SYSTEMS
3/4.6.1 Primary Containment
3/4.6.1.1 Primary Containment Integrity 3.6.1.1 Yes-3 Primary containment functions to mitigate the 3.6.1.3 consequences of a DBA.
5.5.10
3/4.6.1.2 Primary Containment Leakage 3.6.1.1 Yes-3 Primary containment leakage is an assumption utilize d 3.6.1.3 in the LOCA safety analysis (but it is not a process 5.5.10 variable). Therefore. it is being retained to ensure primary containment operability.
3/4.6.1.3 Primary Containment Air Locks 3.6.1.2 Yes-3 Credit for air tightness is considered in safety analysis to limit offsite dose rates during a DBA.
3/4.6.1.5 Primary Containment Structural 3.6.1.1 Yes-3 Primary containment functions to mitigate the Integrity consequences of a DBA.
3/4.6.1.6 Drywell and Suppression Chamber 3.6.1.4 Yes-2 Primary Containment internal pressure is an Internal Pressure assumption of the containment analysis.
3/4.6.1.7 Drywell Average Air Temperature 3.6.1.5 Yes-2 Drywell air temperature is an initial condition in the LOCA safety analysis.
3/4.6.1.8 Drywell And Suppression Chamber 3.6.1.3 Yes-3 Purge isolation valves function to limit DBA Purge System consequences involving offsite release of radioactivity.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.6.2 Depressurization And Cooling Systems
3/4.6.2.1 Suppression Chamber 3.6.2.1 Yes-2,3 Suppression pool water level and temperature are 3.6.2.2 initial conditions in the DBA LOCA analysis and mitigate the consequences of a DBA.
3/4.6.2.2 Suppression Pool Spray 3.6.2.4 Yes-3 Suppression pool spray functions to mitigate the consequences of a DBA LOCA.
3/4.6.2.3 Suppression Pool Cooling 3.6.2.3 Yes-3 Suppression pool cooling functions to limit the consequences of a DBA LOCA.
3/4.6.3 Primary Containment Isolation 3.6.1.3 Yes-3 Isolation valves function to limit DBA consequences.
Valves
3/4.6.4 Vacuum Relief
3/4.6.4.1 Suppression Chamber - Drywell 3.6.1.8 Yes-3 Suppression chamber - drywell vacuum breaker Vacuum Breakers operation is assumed in the LOCA analysis to limit drywell pressure thereby ensuring primary containment integrity.
3/4.6.4.2 Reactor Building - Suppression 3.6.1.7 Yes-3 Reactor building - suppression chamber vacuum Chamber Vacuum Breakers breaker operation is assumed to limit negative pressure differential, secondary to primary containment, that could challenge primary containment integrity.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.6.5 Secondary Containment
3/4.6.5.1 Secondary Containment Integrity 3.6.4.1 Yes-3 Secondary containment limits the offsite dose in accident analyses by ensuring a release from containment is delayed and treated prior to release to the environment.
3/4.6.5.2 Secondary Containment Automatic 3.6.4.2 Yes-3 Valve operation within time limits establishes Isolation Dampers secondary containment and limits offsite dose releases to acceptable values.
3/4.6.5.3 Filtration Recirculation and Ventilation System (FRVS)
3/4.6.5.3.1 FRVS Ventilation Subsystem 3.6.4.3 Yes-3 ISTS 3.6.4.3 Standby Gas Treatment (SGT) System is 5.5.5 renamed Filtration Recirculation and Ventilation System (FRVS), which is the equivalent HCGS specification.
Both the ventilation and recirculation portions of the FRVS are assumed in the DBAs.
5.5.5, Ventilation Filter Test Program (VFTP) based on renumbering 5.5 Specifications.
3/4.6.5.3.2 FRVS Recirculation Subsystem 3.6.4.3 Yes-3 ISTS 3.6.4.3 Standby Gas Treatment (SGT) System is 5.5.5 renamed Filtration Recirculation and Ventilation System (FRVS), which is the equivalent HCGS specification.
Both the ventilation and recirculation portions of the FRVS are assumed in the DBAs.
5.5.5, Ventilation Filter Test Program (VFTP) based on renumbering 5.5 Specifications.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.6.6.2 Drywell and Suppression Chamber 3.6.3.1 Yes-2 Oxygen concentration is limited such that when Oxygen Concentration combined with hydrogen that is postulated to evolve following a LOCA the total concentration remains below explosive levels. Therefore, primary containment integrity is maintained.
ISTS 3.6.3.2 renumbered as ITS 3.6.3.1.
3/4.7 PLANT SYSTEMS
3/4.7.1 Service Water Systems
3/4.7.1.1 Safety Auxiliaries Cooling System 3.7.1 Yes-3 Designed for heat removal for the safety -related systems following a DBA. As such, acts to mitigate the consequences of an accident. The Safety Auxiliaries Cooling System (SACS) is the equivalent of the ISTS RHRSW and DGSW systems. The SACS requirements associated with the RHRSW and DGSW are provided in a single Specification; ITS 3.7.1.
3/4.7.1.2 Station Service Water System 3.7.2 Yes-3 Designed for heat removal for the safety -related systems following a DBA. As such, the Station Service Water System acts to mitigate the consequences of an accident.
3/4.7.1.3 Ultimate Heat Sink 3.7.2 Yes-3 Heat sink for removal of heat from safety -related systems following a DBA. As such, the Ultimate Heat Sink acts to mitigate the consequences of an accident.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.7.2 Control Room Systems
3/4.7.2.1 Control Room Emergency Filtration 3.7.3 Yes-3 Maintains habitability of the control room so that System 5.5.5 operators can remain in the control room following an accident. As such, it mitigates the consequences of an accident by allowing operators to continue accident mitigation activities from the control room.
5.5.5, Ventilation Filter Test Program (VFTP) based on renumbering 5.5 specifications.
3/4.7.2.2 Control Room Air Conditioning (AC) 3.7.4 Yes-3 Maintains habitability of the control room so that System operators can remain in the control room following an accident. As such, It mitigates the consequences of an accident by allowing operators to continue accident mitigation activities from the control room.
3/4.7.4 Reactor Core Isolation Cooling 3.5.3 Yes-3 RCIC is credited in the loss of feedwater transient System analysis to preclude reliance on the low pressure ECCS. Also, retained in accordance with the NRC Final Policy Statement on Technical Specification Improvements due to risk significance.
3/4.7.6 Sealed Source Contamination Relocated No See Appendix A, Page 14.
3/4.7.7 Main Turbine Bypass System 3.7.6 Yes-3 Acts to mitigate the consequences of a feedwater controller failure - minimum demand transient and a turbine trip with bypass event.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.8 ELECTRICAL POWER SYSTEMS
3/4.8.1 A.C. Sources
3/4.8.1.1 AC Sources - Operating 3.8.1 Yes-3 Functions to mitigate the consequences of a DBA.
3.8.3
3/4.8.1.2 AC Sources - Shutdown 3.8.2 Yes-3 Functions to mitigate the consequences of a vessel 3.8.3 draindown event and is needed to support NRC Final Policy Statement requirement for decay heat removal.
3/4.8. 2 D.C. Sources
3/4.8.2.1 D.C. Sources - Operating 3.8.4 Yes -3 Functions to mitigate the consequences of a DBA.
3.8.6
3/4.8.2.2 D.C. Sources - Shutdown 3.8.5 Yes-3 Functions to mitigate the consequences of a vessel 3.8.6 draindown event and is being retained to support the NRC Final Policy Statement requirement for decay heat removal.
3/4.8.3 Onsite Power Distribution Systems
3/4.8.3.1 Distribution - Operating 3.8.7 Yes -3 Functions to mitigate the consequences of a DBA.
3.8.9
3/4.8.3.2 Distribution - Shutdown 3.8.8 Yes-3 Functions to mitigate the consequences of a vessel 3.8.10 draindown event and is being retained to support the NRC Final Policy Statement requirement for decay heat removal.
(a) 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 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.8.4 Electrical Equipment Protective Devices
3/4.8.4.1 Primary Containment Penetration Relocated No See Appendix A, Page 15.
Conductor Overcurrent Protective Devices
3/4.8.4.2 Motor Operated Valves - Thermal Relocated No See Appendix A, Page 16.
Overload Protection (Bypassed)
3/4.8.4.3 Motor Operated Valves - Thermal Relocated No See Appendix A, Page 17.
Overload Protection (Not Bypassed)
3/4.8.4.4 Reactor Protection System Electrical 3.3.8.2 Yes-3 Provides protection for the RPS bus powered Power Monitoring instrumentation against unacceptable voltage and frequency conditions that could degrade the instrumentation so that it would not perform the intended safety function.
3/4.8.4.5 Class 1E Isolation Breaker Relocated No See Appendix A, Page 18.
Overcurrent Protective Devices
3/4.8.4.6 Power Range Neutron Monitoring 3.3.8.3 Yes-3 Provides protection for the PRNM bus powered System Electrical Power Monitoring instrumentation against unacceptable voltage and frequency conditions that could degrade the instrumentation so that it would not perform the intended safety function.
New Specification added: ITS 3.3.8.3.
(a) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 19 of 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.9 REFUELING OPERATIONS
3/4.9.1 Reactor Mode Switch 3.9.1 Yes-3 Provides an interlock to preclude fuel loading with 3.9.2 control rods withdrawn. Operation is assumed in the control rod removal error during refueling and fuel assembly insertion error during refueling accident analysis.
3/4.9.2 Instrumentation 3.3.1.2 Yes - NA Does not satisfy the selection criteria, but is retained because the NRC considers it necessary for flux monitoring during shutdo w n, startup, and refueling operations.
3/4.9.3 Control Rod Position 3.9.3 Yes-3 All control rods are required to be fully inserted when loading fuel. This requirement is assumed as an initial condition in the fuel assembly insertion error during refueling accident analysis.
3/4.9.8 Water Level - Reactor Vessel 3.9. 6 Yes -2 A minimum amount of water is required to assure adequate scrubbing of fission products following a fuel 3/4.9.9 Water Level - Spent Fuel Storage 3.7.7 Yes-2 handling accident.
Pool ISTS 3.7.8 renumbered as ITS 3.7.7.
3/4.9.10.1 Control Rod Removal - Single 3.10.4 Yes/NA The Specification is provided to allow relaxation of Control Rod Removal 3.10.5 certain Limiting Conditions for Operation (LCO) under certain specific conditions to allow testing and 3/4.9.10.2 Control Rod Removal - Multiple 3.10.6 Yes/NA maintenance. The Specification is related to one or Control Rod Removal more LCOs. Direct application of the Technical Specification selection criteria is not appropriate.
(a) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 20 of 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.9.11.1 Residual Heat Removal and Coolant 3.9.7 Yes-4 Retained in accordance with the NRC Final Policy Circulation - High Water Level Statement on Technical Specification Improvements due to risk significance.
ISTS 3.9.8 renumbered as ITS 3.9.7.
3/4.9.11.2 Residual Heat Removal and Coolant 3.9.8 Yes-4 Retained in accordance with the NRC Final Policy Circulation - Low Water Level Statement on Technical Specification Improvements due to risk significance.
ISTS 3.9.9 renumbered as ITS 3.9.8.
3/4.10 SPECIAL TEST EXCEPTIONS
3/4.10.1 Primary Containment Integrity Deleted No Specification is deleted. This Special Test Exception is no longer required at HCGS.
3/4.10.2 Rod Worth Minimizer 3.10. 7 Yes The Specification is provided to allow relaxation of certain Limiting Conditions for Operation (LCO) under 3/4.10.3 Shutdown Margin Demonstrations 3.10.8 Yes certain specific conditions to allow testing and maintenance. The Specification is related to one or more LCOs. Direct application of the Technical Specification selection criteria is not appropriate.
3/4.10.4 Recirculation Loops Deleted No Specification is deleted. This Special Test Exception is no longer required at HCGS.
3/4.10.6 Training Startups Deleted No Specification is deleted. This Special Test Exception is no longer required at HCGS.
(a) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 21 of 22
SUMMARY
DISPOSITION MATRIX FOR HOPE CREEK GENERATING STATION
Current TS (CTS) New TS (ITS) Retained/
Number CTS Specification Title Reference Criterion for Notes Inclusion (a)
3/4.10.8 Inservice Leak and Hydrostatic 3.10.1 Yes The Specification is provided to allow relaxation of Testing certain Limiting Conditions for Operation (LCO) under certain specific conditions to allow testing and maintenance. The Specification is related to one or more LCOs. Direct application of the Technical Specification selection criteria is not appropriate.
3/4.11 RADIOACTIVE EFFLUENTS
3/4.11.1 Liquid Effluents
3/4.11.1.4 Liquid Holdup Tanks 5.5.6 Yes Although this Specification does not meet any criteria of the NRC Final Policy Statement, i t has been retained in accordance with the NRC letter from W. T. Russell to the industry ITS Chairpersons, dated October 25, 1993.
3/4.11.2 Gaseous Effluents
3/4.11.2.7 Main Condenser 3.7.5 Yes-2 The main condenser offgas gross gamma activity rate is an initial safety condition of the Main Condenser Offgas System failure event.
5.0 DESIGN FEATURES 4.0 Yes Application of Technical Specification selection criteria is not appropriate. 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. Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36.
(a) The Applicable Safety Analysis section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.
Page 22 of 22 APPENDIX A
JUSTIFICATION FOR SPECIFICATION RELOCATION
APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.2, Isolation Actuation Instrumentation Table 3.3.2 -1, Function 3.b, Main Steam Line Isolation on Main Steam Line Radiation - High High
DISCUSSION:
The function of the Main Steam Line Radiation Monitoring instrumentation is to measure the radiation levels external to the main steam lines and provide alarm and trip functions upon detection of excessive levels. The main steam line isolation (MSLI) actuation on a Main Steam Line Radiation - High High was originally intended to limit the dose consequences resulting from a radiological release via the Main Steam System during a design basis accident (DBA). This Function initiates closure of the reactor recirculation water sample line isolation valves and does not initiate closure of the main steam isolation valves (MSIVs). Current accident radiological dose analyses assume unfiltered radiological release via the Main Steam System during a DBA until MSIV closure occurs upon receipt of a Reactor Vessel Water Level - Low Low Low, Level 1. The Main Steam Line Radiation - High High Function acts to i solate the reactor recirculation water sample line but does not function to isolate the main steam lines. In addition, the Reactor Coolant System (RCS) sample fluid is maintained inside the reactor building. The reactor building atmosphere is processed via the Filtration Recirculation and Ventilation System (FRVS) to limit offsite radiological releases. Furthermore, any primary containment, MSIV, or e ngineered safety feature component leakage is accounted for in the radiological dose consequences of the accident analyses. In the event that the reactor recirculation water sample line isolation valves are open during a DBA, any radioactivity release via the sample line is terminated by isolation valve closure init iated by a Reactor Vessel Water Level - Low Low, Level 2 signal and does not rely on detection of radiological conditions sensed by the main steam radiation monitors. No DBA or transient takes credit for Main Steam Line Radiation - High High actuation signal.
COMPARISON TO SELECTION CRITERIA:
- 1. The Main Steam Line Radiation - High High instrumentation is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Main Steam Line Radiation - High High instrumentation is not used for monitoring a process variable that is an initial condition of a D BA or transient analysis.
- 3. The Main Steam Line Radiation - High High instrumentation is not used as part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 43) of NEDO -31466, the loss of the Main Steam Line Radiation - High High instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) selection criteria have not been satisfied, the portions of the LCO and Surveillances applicable to the MSLI Main Steam Line Radiation - High High Function may be relocated to other licensee controlled documents outside the Technical Specifications. The Reactor Vessel Water Level - Low Low, Level 2 instrumentation is retained in Technical Specifications to close the reactor recirculation water sample line isolation valves during a DBA or transient requiring RCS sample line
Page 1 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
isolation and the FRVS is retained in the Technical Specifications to filter any radioactive release from the RCS sample system until the sample line is isolated on reactor vessel low water level.
Page 2 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.3, Emergency Core Cooling System Actuation Instrumentation Table 3.3.3-1, Function 4.h, ADS Manual Inhibit Switch
DISCUSSION:
The ADS Manual Inhibit Switch allows the operator to defeat ADS actuation as directed by the emergency operating procedures under conditions for which ADS would not be desirable. For example, during an anticipated transient without scram (ATWS) event low pressure ECCS activation would dilute the sodium pentaborate injected by the Standby Liquid Control (SLC) System thereby reducing the effectiveness of the SLC System ability to shutdown the reactor. 10 CFR 50.62, Requirements for reduction of risk from ATWS events for light-water -cooled nuclear power plants, provid es specific requirements associated with reducing risk from an ATWS event. The alternate rod insertion system, the ATWS recirculation pump trip (RPT) instrumentation and the SLC System provide this risk reduction. As stated in NEDO -31466, the ATWS event is not considered a design basis accident (DBA) or transient and this feature is not required to be retained in the Technical Specifications.
COMPARISON TO SELECTION CRITERIA:
- 1. The ADS Manual Inhibit Switch is not an instrument used for detecting a signi ficant abnormal degradation of the reactor coolant pressure boundary prior to a D BA.
- 2. The ADS Manual Inhibit Switch is not used for, nor capable of, monitoring a process variable that is an initial condition of a D BA or transient analysis.
- 3. The ADS Manual Inhibit Switch is not used as part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier. The inhibit feature allows defeating the automatic ADS function when such action is required by the e mergency operating procedures. However, such manual operator action is not credited in a design basis accident or transient analysis.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 112B) of NEDO-31466, the loss of the ADS Manual Inhibit Switch was found to be a non -significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the portions of the LCO and Surveillances applicable to the ADS Manual Inhibit Switch may be relocated to other licensee controlled documents outside the Technical Specifications. The ATWS RPT instrumentation and the SLC System are retained in Technical Specifications to minimize the risk of an ATWS event pursuant to 10 CFR
- 50. 62.
Page 3 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.6, Control Rod Block Instrumentation Table 3.3.6-1, Function 2, APRM
DISCUSSION:
The Average Power Range Monitor (APRM) control rod block functions to prevent conditions that would require RPS action if allowed to proceed, such as during a "control rod withdrawal error at power." The APRMs utilize LPRM signals to create the APRM rod block signal and provide information about the average core power. However, the APRM rod block function is not used to mitigate a design basis accident (DBA) or transient. The Rod Block Monitor (RBM), Rod Worth Minimizer (RWM), shutdown position of the reactor mode switch, and mode switch refuel position one rod out interlock provide the necessary rod blocks to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a minimum critical power ratio (MCPR ) safety limit (SL) violation, 2) rod withdrawal sequences can effectively limit the potential amount and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, a reactor scram is initiated on high neutron flux from the IRMs and APRMs to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) from a postulated event during a plant startup.
COMPARISON TO SELECTION CRITERIA:
- 1. The APRM control rod block instrumentation is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The APRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysi s.
- 3. The APRM control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 135) of NEDO -3 1466, the loss of the APRM control rod block function was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessmen t.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the APRM control rod block instrumentation LCO and Surveillances applicable to APRM instrumentation may be relocated to other licensee controlled documents outside the Technical Specifications. Rod blocks from the RBM, RWM, and shutdown position of the reactor mode switch, and refuel position one r od out interlock are retained in the Technical Specifications to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a MCPR SL violation, 2) rod withdrawal sequences can effectively limit the potential amount and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, the IRM and APRM Reactor Protection System Functions are retained in the Technical Specifications to preserve the integrity of the fuel cladding and the RCS from a postulated event during a plant startup.
Page 4 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.6, Control Rod Block Instrumentation Table 3.3.6-1, Function 3, Source Range Monitors
DISCUSSION:
Source Range Monitor (SRM) signals are used to monitor neutron flux during refueling, shutdown, and startup conditions. When IRMs are not above Range 2, the SRM control rod block functions to prevent a control rod withdrawal if the count rate exceeds a preset value or falls below a preset limit. No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by the SRMs. The Rod Block Monitor (RBM), Rod Worth Minimizer (RWM), shutdown position of the reactor mode switch, and mode switch refuel position one rod out interlock provide the necessary rod blocks to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a minimum critical power ratio (MCPR) safety limit (SL) violation, 2) rod withdrawal sequences can effectively limit the potential amount and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, a reactor scram is initiated on high neutron flux from the IRMs and APRMs to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) from a postulated event during a plant startup.
COMPARISON TO SELECTION CRITERIA:
- 1. SRM control rod block instrumentation is not used to detect a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The SRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysi s.
- 3. The SRM control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 137) of NEDO -31466, the loss of the SRM control rod block function was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the SRM control rod block instrumentation LCO and Surveillances applicable to SRM instrumentation may be relocated to other licensee controlled documents outside the Technical Specifications. Rod blocks from the RBM, RWM, and shutdown position of the reactor mode switch, and refuel position one rod out interlock are retained in the Technical Specifications to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a MCPR SL violation, 2) rod withdrawal sequences can effectively limit the potential amount and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, the IRM and APRM Reactor Protection System Functions are retained in the Technical Specifications to preserve the integrity of the fuel cladding and the RCS from a postulated event during a plant startup.
Page 5 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.6, Control Rod Block Instrumentation Table 3.3.6-1, Function 4, Intermediate Range Monitors
DISCUSSION:
Intermediate Range Monitors (IRMs) are provided to monitor the neutron flux levels during refueling, shutdown, and startup conditions. The IRM control rod block functions to prevent a control rod withdrawal if the IRM reading exceeds a preset value, or if the IRM is inoperable. No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by IRMs. The Rod Block Monitor (RBM), Rod Worth Minimizer (RWM), shutdown position of the reactor mode switch, and mode switch refuel position one rod out interlock prov ide the necessary rod blocks to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a minimum critical power ratio (MCPR) safety limit (SL) violation, 2) rod withdrawal sequences can effectively limit the potential amou nt and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, a reactor scram is initiated on high neutron flux from the IRMs and APRMs to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) from a postulated event during a plant startup.
COMPARISON TO SELECTION CRITERIA:
- 1. The IRM control rod block instrumentation is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The IRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysi s.
- 3. The IRM control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 138) of NEDO -31466, the loss of the IRM control rod block function was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the IRM control rod block instrumentation LCO and Surveillances applicable to IRM instrumentation may be relocated to other licensee controlled documents outside the Technical Specifications. Rod blocks from the RBM, RWM, and shutdown position of the reactor mode switch, and refuel position one rod out interlock are retained in the Technical Specifications to ensure; 1) control rod withdrawal is limited if localized neutron flux exceeds a predetermined value to preclude a MCPR SL violation, 2) rod withdrawal sequences can effectively limit the potential amount and rate of reactivity increase during a Control Rod Drop Accident, and 3) inadvertent criticality is prevented by limiting control rod withdrawal to one rod during shutdown conditions, respectively. Additionally, the IRM and APRM Reactor Protection System Functions are retained in the Technical Specifications to preserve the integrity of the fuel cladding and the RCS from a postulated event during a plant startup.
Page 6 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.6, Control Rod Block Instrumentation Table 3.3.6-1, Function 5, Scram Discharge Volume
DISCUSSION:
The Scram Discharge Volume (SDV) control rod block functions to prevent control rod withdrawals, utilizing SDV signals to create the rod block signal if water is accumulating in the SDV. The purpose of measuring the SDV water level is to ensure that there is sufficient volume remaining to contain the water discharged by the control rod drives 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 rod withdrawals. With continued water accumulation, a reactor protection system initiated scram signal will occur on a high SDV. Thus, the SDV water level rod block signal provides an opportunity for the operator to take action to avoid a subsequent scram. No DBA or transient takes credit for rod block signals initiated by the SDV instrumentation.
COMPARISON TO SELECTION CRITERIA:
- 1. The SDV control rod block instrumentation is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The SDV control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysi s.
- 3. The SDV control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 139) of NEDO -31466, the loss of the SDV control rod block function was found to be a nonsignificant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Control Rod Block Actuation LCO and Surveillances applicable to SDV instrumentation may be relocated to other licensee controlled documents outside the Technical Specifications. The Reactor Protection System SDV scram function is retained in the Technical Specifications to ensure that there is sufficient SDV remaining to contain the water discharged by the control rod drives during a scram, thus ensuring that the control rods will be able to insert fully.
Page 7 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.7.1, Radiation Monitoring Instrumentation, Function 2, Area Monitors, 2.b. Control Room Direct Radiation Monitor
DISCUSSION:
The Control Room Direct Area Radiation Monitor is used to indicate when the radiation in the control room has exceeded its allowable setpoint. The main control room design is shielded against radiation to allow continued occupancy under accident conditions. Additionally, the Control Room Emergency Filtration (CREF) System and the Control Room Habitability Program requirements provide a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity.
There are no automatic functions that are performed by the Control Room Direct Area Radiation Monitor. The instrument is not used to mitigate a design basis accident (DBA) or transient. Information provided by the Control Room Direct Area Radiation Monitor on the radiation levels within the main control room would have limited u se in identifying / assessing core damage.
COMPARISON TO SELECTION CRITERIA:
- 1. The control room direct area radiation monitor is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The control room direct area radiation monitor does not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The control room direct area radiation monitor does not act as part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 150) of NEDO -31466, the loss of the control room direct area radiation monitor was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Control Room Direct Radiation Area Monitor LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. The main control room design limits radiation exposure and the CREF System and the Control Room Habitability Program requirements are retained in the Technical Specifications to ensure a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity.
Page 8 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.7.1, Radiation Monitoring Instrumentation, Function 2, Area Monitors, 2.a.1) - Criticality Monitors, New Fuel Storage Vault 2.a.2) - Criticality Monitors, Spent Fuel Storage Pool
DISCUSSION:
New fuel storage vault and spent fuel storage pool area criticality monitors are provided to detect excessive radiation levels as an indication of criticality. There are no automatic functions that are performed by these instruments. The instruments are not used to mitigate a design basis accident (DBA) or transient. 10 CFR 70.24 and 10 CFR 50.68 govern c riticality accident requirements.
10 CFR 70.24(d)(1) and 10 CFR 50.68(a) allow for the holder of an operating license to comply with either 10 CFR 70.24(a) through (c) or 10 CFR 50.68(b). Hope Creek complies with 10 CFR 50.68(b) in lieu of maintaining monitoring systems as described in 10 CFR 70.24.The design of the new fuel storage vault and spent fuel storage pool provide a margin to criticality pursuant to 10 CFR 50.68(b) precluding criticality.
COMPARISON TO SELECTION CRITERIA:
- 1. The New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors are not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors do not monitor a process variable that is an initial condition of a DBA or transient analyse s.
- 3. The New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors do not act as part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. Consistent with the discussion provided for area radiation monitors in NEDO -31466 Sections 3.5 and 6, and Table 4-1 (item 150), PSEG has determined that the loss of the New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors is a non-significant risk contributor to Hope Creek core damage frequency and offsite releases. The New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors are not addressed in the Hope Creek 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 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the New Fuel Storage Vault and Spent Fuel Storage Pool Area Criticality Monitors LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. The design of the new fuel storage vault and spent fuel storage pool will continue to provide a margin to criticality pursuant to 10 CFR 50.68(b) precluding criticality.
Page 9 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.7.1, Radiation Monitoring Instrumentation, Function 3, Reactor Auxiliaries Cooling Radiation Monitor
DISCUSSION:
The Reactor Auxiliaries Cooling System (RACS), which provides cooling to the non-safety related components, is monitored continuously at the outlet of the RACS pumps to detect leakage of radioactivity into the systems. If RACS water should become contaminated due to the leakage from one of its components, the Reactor Auxiliaries Cooling Radiation Monitor will alarm in the main control room alerting the operators so the RACS loop can be isolated from the rest of the Station Service Water System. There are no automatic functions that are performed by this instrument. The instrument provides indication only and does not provide meaningful information following a design basis accident (DBA) or transient. Additionally, the Reactor Auxiliaries Cooling Radiation Monitor is not used to mitigate a DBA or transient. The Radioactive Effluents Control Program provides controls to conform with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to member(s) of the public from radioactive effluents as low as reasonably achievable. The R adioactive Effluents Control Program Technical Specification requires the program requirements to be contained in the Offsite Dose Calculation Manual.
COMPARISON TO SELECTION CRITERIA:
- 1. The Reactor Auxiliaries Cooling Radiation Monitor is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Reactor Auxiliaries Cooling Radiation Monitor does not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The Reactor Auxiliaries Cooling Radiation Monitor is not the part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 142) of NEDO -31466, the loss of the Reactor Auxiliaries Cooling Radiation Monitor was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Reactor Auxiliaries Cooling Radiation Monitor LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. Program requirements associated with the Radi oactive Effluents Control Program are retained in the Technical Specifications, providing controls to conform with 10 CFR 50.36a.
Page 10 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.7.1, Radiation Monitoring Instrumentation, Function 4, Safety Auxiliaries Cooling Radiation Monitor
DISCUSSION:
The Safety Auxiliaries Cooling System (SACS) is monitored continuously downstream of the residual heat removal heat exchanger to detect inleakage of radioactivity into the system. In addition, sample points are provided at selected equipment locations to facilitate leak detection. Such leakage accumulates in the SACS expansion tank and eventually causes a high level that alerts the operator to the abnormal condition. There are no automatic functions that are performed by this instrument. The instrument provides indication only and does not provide meaningful information following a design basis accident (DBA) or transient. Additionally, the Safety Auxiliaries Cooling Radiation Monitor is not used to mitigate a DBA or transient. The Radioactive Effluents Control Program provides controls to conform with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to member(s) of the public from radioactive effluents as low as reasonably achievable. The Radioactive Effluents Control Program Technical Specification requires program requirements to be contained in the Offsite Dose Calculation Manual.
COMPARISON TO SELECTION CRITERIA:
- 1. The Safety Auxiliaries Cooling Radiation Monitor is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Safety Auxiliaries Cooling Radiation Monitor does not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The Safety Auxiliaries Cooling Radiation Monitor is not the part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 142) of NEDO -31466, the loss of Safety Auxiliaries Cooling Radiation Monitor was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Safety Auxiliaries Cooling Radiation Monitor LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. Program requirements associated with the Radioactive Effluents Control Program are retained in the Technical Specifications, providing controls to conform with 10 CFR 50.36a.
Page 11 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.3.7.1, Radiation Monitoring Instrumentation, Function 5, Offgas Pre-treatment Radiation Monitor
DISCUSSION:
The Offgas Pre-treatment Radiation Monitor detects the radiation concentration that is attributable to the non-condensable fission product gases that are produced in the reactor and transported with steam through the turbine to the condenser. Changes in the radiation values can be used to interpret fuel rod condition. There are no automatic functions that are performed by this instrument. The instrument provides an alarm function only and does not provide meaningful information following a design basis accident (DBA) or transient. Additionally, the Offgas Pre-treatment Radiation Monitor is not used to mitigate a DBA or transient. Limits on the linear heat generation rate (LHGR), average planar linear heat generation rate (APLHGR), and minimum critical power ratio (MCPR) are specified to ensure that the fuel design limits are not exceeded and no fuel damage results during normal operation and anticipated operational occurrences (AOOs).
COMPARISON TO SELECTION CRITERIA:
- 1. The Offgas Pre-treatment Radiation Monitor is not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Offgas Pre-treatment Radiation Monitor does not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The Offgas Pre-treatment Radiation Monitor is not the part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
- 4. The Offgas Pre-treatment Radiation Monitor is not addressed in the Hope Creek PRA and does not represent a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and s afety.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Offgas Pre-treatment Radiation Monitor LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. Limits on the LHGR, APLHGR, and MCPR are retained in the Technical Specifications to ensure that the fuel d esign limits are not exceeded and no fuel damage results during normal operation and AOOs.
Page 12 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
CTS 3/4.3.7.5, Accident Monitoring Instrumentation Table 3.3.7.5-1, Functions 7, 11, 12, and 13
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 scr eening criteria to post-accident monitoring instrumentation is documented in a letter dated May 9, 1988 from T.E. Murley (NRC) to the Owners Groups (Reference 3). 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 (inc luding long term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accidents (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 shoul d contain Type A instruments and non-Type A Category 1 instruments specified in the plant's s afety evaluation report (SER) on RG 1.97. Accordingly, this position has been applied to the Hope Creek RG 1.97 instruments. The following CTS Table 3.3.7.5-1 Instruments do not meet RG 1.97 Type A or non-Type A, Category 1 requirements :
Function 7 Drywell Air Temperature Function 11 North Plant Vent Radiation Monitor Function 12 South Plant Vent Radiation Monitor Function 13 FRVS Vent Radiation Monitor
COMPARISON TO SELECTION CRITERIA:
- 1. These instruments are not used for 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 Hope Creek 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 10 CFR 50.36(c)(2)(ii) screening 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.
Page 13 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
CTS 3/4.7.6, 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 < 0.00 5 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 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 design basis accident (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. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 267) of NEDO -31466, Sealed Source Contamination was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to HCGS, and concurs with the assessment
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening 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 license of Hope Creek.
Page 14 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.8.4.1, Primary Containment Penetration Conductor Overcurrent Protective Devices
DISCUSSION:
The primary feature of these protective devices is to open the control and/or power circuit whenever the load conditions exceed the present current demands. This is to protect the circuit conductors against damage or failure due to overcurrent heating effects. The con tinuous monitoring of the operating status of the overcurrent protection devices is impracticable and not covered as part of control room monitoring, except after trip condition indication.
In the event of failure of the protective device to trip the circuit, the upstream protective device is expected to operate and isolate the faulty circuit. Thus, the backup protection will prevent loss of the redundant power source. In the worst case fault condition, a single division of protective functions can be lost. However, this scenario is covered under the single failure criterion.
The overcurrent protection devices ensure the pressure integrity of the containment penetration. With failure of the device it is postulated that the wire insulation will degrade resulting in a containment leak path during a LOCA. However, containment leakage is not a process variable and is not considered as part of the primary success path. Containment penetration degradation will be identified during the normal containment leak rate tests required by 10 CFR 50, Appendix J.
COMPARISON TO SELECTION CRITERIA:
- 1. The primary containment penetration conductor overcurrent protective devices are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary.
- 2. The primary containment penetration conductor overcurrent protective devices do not monitor a process variable that is an initial condition of a design basis accident ( DBA) or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The primary containment penetration conductor overcurrent protective devices are not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 276) of NEDO -31466, the loss of the circuits associated with the primary containment penetration conductor overcurrent protective devices was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it applicable to HCGS, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Primary Containment Penetration Conductor Overcurrent Protective Devices LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. Compliance with applicable portions of 10 CFR 50, Appendix J is required by the operating license of Hope Creek.
Page 15 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.8.4.2, Motor Operated Valves - Thermal Overload Protection (Bypassed)
DISCUSSION:
For motor operated valve operators with thermal overload protection bypassed (i.e., bypass the thermal overload trip continuously or only under accident conditions), the valve function should be accomplished even if the overload trip is sensed. The valve function for these valves is meant to take precedence over the overload protection. If the overload condition occurs during valve operation, in case of failure of overload protection bypass circuit, the safety function of the valve will not be performed. This affects the Operability of the system containing the valve. Accordingly, the more appropriate LC O would be to address the overall system Operability and not the Operability of a support system. Additionally, the surveillance and maintenance of the devices can be controlled by sources other than the plant Technical Specifications.
COMPARISON TO SELECTION CRITERIA:
- 1. Motor operated valve thermal overload b ypass protection is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. Motor operated valve thermal overload b ypass protection is not, and does not monitor, a process variable that is an initial condition of a DBA or transient.
- 3. By passing of a motor operated valve's thermal overload protection is not part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier. The supported system (e.g., ECCS) may be part of a success path and is retained in the Technical S pecifications. However, motor operated valve thermal overload bypass protection retention in the Technical Specifications is not necessary as its safety related role is implicitly addressed in the Operability determination of the supported system.
- 4. PRAs address system risk contribution and identify systems which can be significant risk contributors to core damage and offsite releases. Subcomponent risk contribution from motor operated valve thermal overload b ypass protection malfunction is not addressed as the inoperabil ity of the supported system and subsequent risk contribution is inclusive of thermal overload protection failure potential.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Motor -Operated V alves -
Thermal Overload Protection (Bypassed) LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. The Specifications retained in the Technical Specifications continue to require motor operated valves assumed to operate during a DBA or transient to be capable of performing their intended safety functions.
Page 16 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.8.4.3, Motor Operated Valves - Thermal Overload Protection (Not Bypassed)
DISCUSSION:
For motor operated valve operators with thermal overload protection (i.e., trip on overload condition), the valve function should be accomplished even if the overload trip is sensed. The valve function for these valves is meant to take precedence over the overload protection. If the overload condition occurs during valve operation, in case of failure of overload protection operation to disconnect the load, the safety function of the valve will not be performed. This affects the Operability of the system containing the valve. Accordingly, the more appropriate LC O would be to address the overall system Operability and not the Operability of a support system. Additionally, the surveillance and maintenance of the devices can be controlled by sources other than the plant Technical Specifications.
COMPARISON TO SELECTION CRITERIA:
- 1. Motor operated valve thermal overload protection is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. Motor operated valve thermal overload protection is not, and does not monitor, a process variable that is an initial condition of a DBA or transient.
- 3. Actuation of a motor operated valve's thermal overload protection is not part of a primary success path in the mitigation of a DBA or transient that assumes the failure of, or presents a challenge to the integrity of a fission product barrier. The supported system (e.g., ECCS) may be part of a success path and is retained in the Technical S pecifications. However, motor operated valve thermal overload protection retention in the Technical Specifications is not necessary as its safety related role is implicitly addressed in the Operability determination of the supported system.
- 4. PRAs address system risk contribution and identify systems which can be significant risk contributors to core damage and offsite releases. Subcomponent risk contribution from motor operated valve thermal overload protection malfunction is not addressed as the inoperabil ity of the supported system and subsequent risk contribution is inclusive of thermal overload protection failure potential.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Motor -Operated V alves -
Thermal Overload Protection (Not bypassed) LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. The Specifications retained in the Technical Specifications continue to require motor operated valves assumed to operate during a DBA or transient to be capable of performing their intended safety functions.
Page 17 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
3/4.8.4.5, Class 1E Isolation Breaker Overcurrent Protective Devices
DISCUSSION:
The Class 1E isolation breaker overcurrent protective devices sense and protect the Class 1E buses from overcurrent conditions of non-Class 1E electrical loads during operational and shutdown modes by opening the load supply breaker isolating the load from the Class 1E buses. The Class 1E onsite AC sources and the offsite power sources and their distribution system are of sufficient capacity and capability to supply power to both Class 1E and non-Class 1E loads during plant conditions, including design basis accidents (DBAs) and transients. In the event of a LOCA, the non-Class 1E loads are automatically tripped from the Class 1E buses. Therefore, failure of one or more non-Class 1E electrical load overcurrent protection devices during a DBA or transient would not prevent Class 1E loads from performing their intended safety functions. The Class 1E AC Electrical Power System design supports the loss of a single Class 1E AC electrical power distribution subsystem (i.e., channel), which bounds an electrical fault on a non-Class 1E load with no overcurrent protection. Technical Specifications require four AC electrical power distribution subsystems to be OPERABLE and the failure of a single AC electrical subsystem is considered in the Limiting Condition for Operation (LCO) of the Class 1E AC Electrical Power Distribution System. Also, maintaining the quality of the Class 1E AC Electrical Power System, which includes the non-Class 1E breaker overcurrent protective devices and the capability to isolate non-Class 1E loads from the Class 1E loads in the event of an overcurrent condition, is required pursuant to 10 CFR 50, Appendix B.
COMPARISON TO SELECTION CRITERIA:
- 1. The Class 1E Isolation Breaker Overcurrent Protective Devices are not used for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Class 1E Isolation Breaker Overcurrent Protective Devices are not, and do not monitor, a process variable, and do not represent a design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrie r.
- 3. The Class 1E Isolation Breaker Overcurrent Protective Devices are not structures, systems, or components that comprise the primary success path or 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. As discussed in Sections 3.5 of NEDO-31466, and summarized in Section 5 and Table 6-1 (item 375) of Supplement 1 to NED O-31466, the failure of the Class 1E Isolation Breaker Overcurrent Protective Devices to isolate non-safety related loads during DBAs and transient events other than a LOCA was found to be a non-significant risk contributor to core damage frequency and offsite releases. PSEG has reviewed this evaluation, considers it appl icable to Hope Creek, and concurs with the assessment.
CONCLUSION:
Since the 10 CFR 50.36(c)(2)(ii) screening criteria have not been satisfied, the Class 1E Isolation Breaker Overcurrent Protective Devices LCO and Surveillances may be relocated to other licensee controlled documents outside the Technical Specifications. Operability of the four AC electrical power
Page 18 of 19 APPLICATION OF SELECTION CRITERIA TO THE HOPE CREEK GENERATING STATION TECHNICAL SPECIFICATIONS
Appendix A - Justification for Specification Relocation
distribution subsystems is retained in the Technical Specifications to ensure electrical power is provided to equipment assumed to operate during a DBA or transient. Compliance with applicable portions of 10 CFR 50, Appendix B related to maintaining quality of equipment to an extent consistent with their importance to safety is required by the operating license of Hope Creek.
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