ML25189A154
| ML25189A154 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/08/2025 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Jurek S Public Service Enterprise Group |
| References | |
| EPID L-2025-LLA-0021, EPID L-2025-LLA-0022 | |
| Download: ML25189A154 (1) | |
Text
{{#Wiki_filter:From: Richard Guzman To: Jurek, Shane
Subject:
Salem Generating Station, Units 1 and 2 - Risk-Informed Completion Time TSTF-505/TSTF-591 and 10 CFR 50.69 (2nd Set) Audit Questions (EPIDs L-2025-LLA-0021, L-2025-LLA-0022) Date: Tuesday, July 8, 2025 9:11:56 AM Attachments: Salem Audit Questions (2nd set) 7-8-25.pdf
- Shane, By letters dated January 31, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML25031A416 and ML25031A371), PSEG Nuclear, LLC (PSEG, the licensee) submitted two license amendment requests (LARs) for Salem Generating Station, Units 1 and 2 (Salem). The proposed amendments would modify Salem licenses DPR-70 and DPR-75 and the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler 505 (TSTF-505), Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, TSTF-591, Revise the Risk Informed Completion Time (RICT) Program, and the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.
On April 14, 2025, the NRC staff issued an audit plan (ML25096A002) that conveyed intent to conduct a regulatory audit to support its review of the subject licensing actions. Based on the commonalities between the LARs and subsequent overlap in technical content and review personnel, the staff is conducting a combined audit that addresses both LARs. In the audit plan, the NRC staff requested an electronic portal setup and provided a list of documents to be added to the online portal. The audit plan also indicated that the NRC may request information and audit meetings/interviews throughout the audit period. The NRC staff performed an initial review of the list of documents and has developed a list of audit questions. The first set of audit questions were sent to you via email on May 8, 2025 (ML25169A316) and the associated audit meeting with the NRC staff is scheduled for July 15th. The second set of questions are provided in the attachment. Please post the response(s) for the questions to the online portal as the responses are completed. The staff would like to target the end of July for a virtual audit call to discuss the responses to the second set of audit questions, if possible. Please contact me at any time prior if a clarification discussion is needed. We look forward to discussing these questions and PSEGs responses during the virtual audit meeting. Thank you, Rich Guzman Sr. PM, Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Phone: (301) 415-1030 Richard.Guzman@nrc.gov
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AUDIT QUESTIONS LICENSE AMENDMENT REQUESTS TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2, RISK-INFORMED EXTENDED COMPLETION TIMES INITIATIVE 4B; TSTF-591, REVISION 0, RISK-INFORMED COMPLETION TIME PROGRAM; AND 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS PSEG NUCLEAR, LLC SALEM GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-272 AND 50-311 By letters dated January 31, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML25031A416 and ML25031A371), PSEG Nuclear LLC submitted license amendment requests (LARs) to amend the licenses for Salem Generating Station, Units 1 and 2, Renewed Facility Operating License Nos. DPR-70 and DPR-75, respectively. The proposed LARs would adopt Technical Specifications Task Force Traveler 505 (TSTF-505), Revision 2, Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, TSTF-591, Revise the Risk Informed Completion Time (RICT) Program, and the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. On April 14, 2025, the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan (ML25096A002) that conveyed intent to conduct a regulatory audit to support its review of the subject LARs. Based on the commonalities between the LARs and subsequent overlap in technical content and review personnel, the NRC staff is conducting a combined audit that addresses both LARs. The NRC staff has determined that additional information is needed to support the audit as shown in the following audit questions. The questions are ordered by number and identified by technical review branch/area. Audit Question-14 (APLA-01), Common Cause Failure (CCF) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A, states that no common cause failure (CCF) adjustment is required for planned maintenance. The NRC safety evaluation (SE) for NEI 06-09, Revision 0, dated May 17, 2007 (ML071200238), is based on conformance with Regulatory Guide (RG) 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, dated May 2011 (ML100910008). Specifically, the NRC SE, Section 2.2, states that, specific methods and guidelines acceptable to the NRC staff are [] outlined in RG 1.177 for assessing risk-informed TS changes. The NRC SE, Section 3.2, further states that compliance with the guidance of RG 1.174, Revision 1, An Approach for Using Probabilistic
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated November 2002 (ML023240437), and RG 1.177, Revision 1, is achieved by evaluation using a comprehensive risk analysis, which assesses the configuration-specific risk by including contributions from human errors and common cause failures. RG 1.177, Revision 2, dated January 2021 (ML19206A493) provides one acceptable approach to addressing the guidelines in RG 1.177, Revision 1. RG 1.177, Revision 2, Section 2.3.3.1, states that, CCF modeling of components is not simply dependent on the number of remaining inservice components; it is also dependent on the reason the components were removed from service (i.e., whether for preventative or corrective maintenance). In relation to CCF for preventive maintenance, the guidance in RG 1.177, Appendix A, Section A-1.3.1.1, states: If the component is down because it is being brought down for maintenance (but not failed), the CCF contributions involving the component should be modified to remove the component and to only include failures of the remaining components (also see Section C.2.3.3 of this RG). According to the guidance of RG 1.177, Revision 2, if a component from a CCF group of three or more components is declared inoperable, the CCF of the remaining components should be modified to reflect the reduced number of available components in order to properly model the as-operated plant. The Salem TSTF-505 LAR addresses how CCFs are treated in the PRA models for planned maintenance in Attachment 1, Insert 5 for Section 6.19d. Therefore, address the following: a) Explain how CCF is included in the PRA model (e.g., with all combinations in the logic models as different basic events or with identification of multiple basic events in the cut sets). b) Describe or demonstrate, using the real-time risk Phoenix model, how the treatment of CCF is evaluated for configuration-specific risk when quantifying a risk-informed completion time (RICT). Explain how the quantification and/or models will be changed when, for example, one train of a 3x100 percent train system is removed for preventative maintenance vs. emergent or corrective maintenance. Audit Question-15 (APLA-02), FLEX Mitigating Capability NRC memorandum dated May 6, 20221 provides the NRCs staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.2002. LAR Enclosure 9, Table E9-1, Assessment of Internal Events PRA Epistemic Uncertainty Impacts outlines that HRA is considered and that LCOs related to offsite or onsite AC sources have an effect on the RICT. 1 U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments, dated May 6, 2022 (ML22014A084). 2 U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk Informed Activities, RG 1.200, Revision 3, December 2020 (ML20238B871).
a) Explain how FLEX is addressed in the model and is considered for RICT estimates. b) Discuss the function and relationship between the Supplemental Power System (SPS) and offsite/onsite emergency power. Audit Question-16 (APLA-03), State of Knowledge Correlation (SOKC) LAR Enclosure 9, Table E9-1, Assessment of Internal Events PRA Epistemic Uncertainty Impacts, notebooks SA-LAR-025, Assessment of Key Assumptions and Sources of Uncertainty for the Salem Generating Station PRA and SA-PRA-018, Uncertainty Notebook, describe how key assumptions and sources of uncertainty are identified for the TSTF-505 program and how RICTs could be impacted. ASME/ANS Standard supporting requirements QU-A3 and QU-E3, EPRI TR 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, and Section 6.4 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, dated March 2017 (ML17062A466) address specific guidance for SOKC. For a CC II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to are the means of the probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly reflected in the PRA models. In general, the point estimate CDF and LERF obtained by quantification of the cutset probabilities using mean values for each basic event probability does not produce a true mean of the CDF/LERF. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the SOKC is unimportant (i.e., the risk results are well below the acceptance guidelines). a) Discuss the impact of SOKC to the Salem model, if applicable to the baseline CDF and LERF estimates. b) Discuss and explain any need or applicability of SOKC to RICT estimates. Audit Question-17 (APLA-04), Configuration Alignment and Shared Systems Shared systems and alignments are described in the LAR Enclosure 8, Attributes of the Real Time Risk Model. a) Clarify and demonstrate how alternate system alignments and system asymmetries, such as CREACS, control air, and service air are considered in the PRA and real-time quantification models, i.e., Phoenix, for applicable systems and provide justification that these treatments do not underestimate CDF and LERF. b) If CDF and LERF are determined to be underestimated, then justify that this non-conservatism is inconsequential to the RICT estimates. Audit Question-18 (APLA-05), Power-Operated Relief Valves (PORV) Block Valves LAR Table E1-1, In Scope TS/LCO Conditions to Corresponding PRA Functions outline the success criteria for PORVs and their corresponding relief valves in LCO 3.4.3 (Unit 1) and LCO
3.4.5 (Unit 2). There are two PORV and Block Valve success criteria for feed-and-bleed, dependent on injection source (charging pump vs. safety injection pump). a) Discuss or demonstrate how the various success criteria are accounted for in the model to address conditions for use, i.e., application of feed-and-bleed following a loss of AFW? b) Additionally, plants have experienced leaky PORVs in the past, which required long-term block valve closure as allowed by Technical Specifications. If leaky PORVs are modeled in the Salem PRA, explain how the potential for leaky PORVs has impacted the model and discuss or demonstrate how configurations such as this are addressed in the real-time Phoenix risk model for RICT estimates. Audit Question-19 (APLA-06), Digital I&C NEI 06-09 states regarding the quality of the PRA model that: RG 1.174, Revision 1, and RG 1.200, Revision 1 define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. Regarding digital I&C, NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed applications. In addition, known modeling challenges exist, such as the lack of industry data for digital I&C components, differences between digital and analog system failure modes, and the complexities associated with modeling software failures, including common cause software failures. Also, though reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program. However, it is not clear to NRC staff whether the licensee credited digital systems in the PRA models that will be used in the RICT program or whether this modelling can impact the RICT calculations. Therefore, address the following: a) Clarify whether digital I&C systems are credited in the PRA models that will be used in the RICT program. b) If digital I&C systems are credited in the PRA models that will be used in the RICT program, then justify that the modeling uncertainty associated with crediting digital I&C systems in the PRA models is appropriate. Describe its impact on RICT calculations. Audit Question-20 (APLA-07), Uncertainty On review of SA-PRA-018, PRA Uncertainty Notebook, Table A-1, Representative Issue Characterization for Sources of Model Uncertainty (QU-F4 and LE-F3), it was noted that Topic 9 identified room heat up calculation uncertainty as part of a compensatory measure. a) Discuss or demonstrate if a sensitivity calculation had been done for this measure? If not, justify whether it should be done. b) For the action of using portable fans during SBO conditions, is a safety-related safeguards (emergency)-derived power source credited? If not, describe and justify the effectiveness of this compensatory measure using non-safety related power sources on the RICT estimates.
Audit Question-21 (APLB-01), In reviewing the LAR Enclosure 5, Table E5-1, the NRC staff noted that fire contribution to the baseline CDF/LERF is over 90% for Salem Units 1 & 2. The NRC staff requests clarification on the factors that resulted in fire events being such a highly dominant contribution to CDF/LERF. Audit Question-22 (APLB-02) The key factors used to justify using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, "Recent Fire PRA Methods Review Panel Decisions and Electrical Power Research Institute (EPRI) 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires'." (ML12172A406). If any reduced transient HRRs below the bounding 98% HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion: a) Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied. b) A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location. c) The results of a review of records related to compliance with the transient combustible and hot work controls. Audit Question-23 (APLB-03) FAQ 13-0004, "Clarifications on Treatment of Sensitive Electronics" (ML13322A085) provides supplemental guidance for application of the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850, Volume 2, for solid-state and sensitive electronics. a) Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents). b) If the approach cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no impact on the RICT calculations. c) As an alternative to item b above, add an implementation item to replace the current approach with an acceptable approach prior to the implementation of the RICT program. Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance. Audit Question-24 (APLB-04)
NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines-Final Report," (ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in human reliability analyses (HRAs). NUREG-1921 refers to Table 2-1 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA)," (ML051160213), which recommends that joint human error probability (HEP) values should not be below 1E-5. Table 4-4 of EPRI 1021081, "Establishing Minimum Acceptable Values for Probabilities of Human Failure Events," provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency. LAR Enclosure 9, Table E9-3 stated that the HEPs used in the FPRA were adjusted to consider the additional challenges that may be present given a fire. It is unclear to the NRC staff whether a value less than 1E-5 was used in the FPRA, and if so, address the following: a) Provide justification, such as a RICT sensitivity study that the that the minimum joint HEP value has no impact on the remaining TS LCOs proposed for the RICT application. b) If, in response part (a), if it cannot be justified that the minimum joint HEP value has no impact on the application, then provide the following:
- i.
Confirm that each joint HEP value used in the FPRA below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline (i.e., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence). Provide an estimate of the number of these joint HEP values below 1.0E-5, discuss the range of values, and provide at least two different examples where this justification is applied. ii. If joint HEP values used in the FPRA below 1E-5 cannot be justified, add an implementation item to set these joint HEPs to 1E-5 in the FPRA prior to the implementation of the RICT program. Audit Question-25 (APLB-05) NUREG-2178, Volume 1 "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ML16110A14016) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction. Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet. a) If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet. b) Justify any modelling in which the base of an obstructed plume is located at less than one half of the cabinet's height. c) As an alternative to item b above, add an implementation item to remove credit for the obstructed plume model in the FPRA prior to the implementation of the RICT program.
Audit Question-26 (APLB-06) Guidance in FAQ 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of "robustly or well-sealed." Thus, for cabinets of 440V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440V and higher, the original guidance in Chapter 6 remains and states that Bin 15 panels which house circuit voltages of 440V or greater are counted because an arcing fault could compromise panel integrity (an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires)." FPRA FAQ 14-0009, "Treatment of Well-Sealed MCC Electrical Panels Greater than 440V" (ML15119A176) provides the technique for evaluating fire damage from MCC cabinets having a voltage greater than 440V. The NRC guidance recommends that propagation of fire outside the ignition source panel should be evaluated for all MCC cabinets that house circuits of 440V or greater. a) Describe how fire propagation outside of well-sealed MCC cabinets greater than 440V is evaluated. b) If well-sealed cabinets less than 440V are included in the Bin 15 count of ignition sources, provide justification for using this approach as this is contrary to the guidance. Audit Question-27 (APLB-07) NUREG/CR-6850, Section 6, "Fire Ignition Frequencies," and FAQ 12-0064 "Hot Work/Transient Fire Frequency Influence Factors" (ML12346A488) describes the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance: a) Indicate whether the methodology used to calculate hot work and transient fire frequencies applies influencing factors using NUREG/CR-6850 guidance or FAQ 12-0064 guidance. b) Indicate whether administrative controls are used to reduce transient fire frequency, and if so, describe and justify these controls. c) Indicate whether you have any combustible control violations and discuss your treatment of these violations for the assignment of transient fire frequency influence factors. For those cases where you have violations and have assigned an influence factor of 1 (Low) or less, indicate the value of the influence factors you have assigned and provide your justification. d) If you have assigned an influencing factor of "0" to Maintenance, Occupancy, or Storage, or Hot Work for any fire physical analysis units (PAUs) provide justification. e) If a weighting factor of "50" was not used in any fire PAU, provide a sensitivity study that assigns weighting factors of "50" per the guidance in FAQ 12-0064. Audit Question-28 (APLB-08) The licensee indicated that Salem Units 1 and 2 has shared systems between units. Therefore, address the following:
a) Explain how the risk contribution of fires originating in one unit is addressed for the other unit given impacts due to the physical proximity of equipment and cables in one unit to equipment and cables in the other unit. Include identification of locations where fire in one unit can affect components in the other unit and explain how the risk contributions of such scenarios are allocated in the LAR. b) Explain how the contributions of fires in common areas are addressed, including the risk contribution of fires that can impact components in both units. c) Explain the extent to which systems are shared by both units and whether shared systems are credited in the PRA models for both units. If shared systems are credited in the PRA models for each unit, then explain how the PRAs address the possibility that a shared system is demanded in both units in response to a single initiating event or fire initiator. Audit Question-29 (APLB-09) Traditionally, the cabinets on the front face of the main control board (MCB) have been referred to as the MCB for purposes of FPRA. Appendix L of NUREG/CR-6850, (ML052580075) provides a refined approach for developing and evaluating those fire scenarios. FPRA FAQ 14-0008, "Main Control Board Treatment" dated July 22, 2014 (ML14190B307) clarifies the definition of the MCB and provides guidance for when to include the cabinets on the back side of the MCB as part of the MCB for FPRA. It is important to distinguish between MCB and non-MCB cabinets because misinterpretation of the configuration of these cabinets can lead to incomplete or incorrect fire scenario development. This FAQ also provides several alternatives to NUREG/CR-6850 for using Appendix L to treat partitions in an MCB enclosure. Therefore, address the following: a) Describe the main control room MCB configuration, and use the guidance in FAQ 14-0008, to determine whether cabinets on the rear side of the MCB are a part of the MCB. Provide justification using the FAQ guidance. b) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure using the definition in FAQ 14-0008 confirm that guidance in FAQ 14-0008 was used to develop fire scenarios in the MCB and determine the frequency of those scenarios. c) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure and the guidance in FAQ 14-0008 was not used to develop fire scenarios involving the MCB provide a description of how the fire scenarios for the backside cabinets are developed and an explanation of how the treatment aligns with NRC accepted guidance. d) If in response to parts (c) above, the current treatment of the MCB and those cabinets on the rear side of the MCB cannot be justified using NRC accepted guidance, then justify that the treatment has no impact on the RICT calculations. Alternatively, propose a mechanism that ensures that the FPRA is updated to treat the MCB enclosure consistent with the guidance in FAQ 14-0008, prior to implementation of the RICT program. Audit Question-30 (APLC-01) Both 10 CFR 50.69 and TSTF-505 programs are approved and implemented based on a technically adequate modeling of risk. Although hazards do not need to be treated identically in
both programs, conservative assumptions are relied on if a PRA model of sufficient technical adequacy is not used to address a hazard group. Section 2.3.1, item 7, of NEI 06-09-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs safety evaluation for NEI 06-09 states that where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT. Another method to address external event risks that are not modeled in the PRA is to demonstrate that the external event is not a significant contributor to risk. The safety evaluation for NEI 06-09 states that other external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration specific risk. The Salem 10 CFR 50.69 application credits a list of doors shown in attachment 1 of operating procedure SC.OP-AB.ZZ-0001 to screen local intense precipitation (LIP), hurricane, storm surge, and external flood hazards. The Salem 10 CFR 50.69 application also states that these doors shall not be categorized any lower than HSS, as they are credited for screening this hazard in accordance with figure 5-6 of NEI 00-04. The Salem TSTF-505 application also screened these hazards crediting the list of doors shown in attachment 1 of operating procedure SC.OP-AB.ZZ-0001. These hazards, except for LIP, were screened based on preliminary screening criterion C5, Event develops slowly, allowing adequate time to eliminate or mitigate the threat. A basis is provided that states, Criterion C5 also applies to the screening of the PMHSS [Probable Maximum Hurricane Storm Surge], as the station will be in a shutdown condition well before the arrival of flood waters that could challenge the CDB [current design basis] flood protection features. Although PSEG procedure OP-AA-111-1001, revision 17, directs operators to place the units in a shutdown condition well before the arrival of a hurricane and adverse weather conditions, it effectively removes the station from the RICT program as it only applies to units while at power. Clarify why the risk this external hazard poses to the plant does not need to be considered while still in a mode of operation where the RICT program is active. Audit Question-31 (APLC-02) Section 2.3.1, item 7, of NEI 06-09 states that the impact of other external events risk shall be addressed in the Risk-Managed Technical Specifications (RMTS) program. It explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated Risk-Informed Completion Time (RICT). The NRC staffs safety evaluation for NEI 06-09 states that where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT. Section 5.5.2 of enclosure 4 of the TSTF-505 License Amendment Request (LAR) states that the PMHSS and combined flooding mechanism is bound by the current design basis (CDB) of the plant and that the site has adequate flood protection features. It further states that (1) the service water intake structure (SWIS) and power block buildings are protected from flood waters up to a design elevation of 113.8 feet PSD with waves affecting the power block to 120.4 feet and the SWIS building to 127.3 feet PSD and (2) the Flood Hazards Reevaluation Report (FHRR) calculated a maximum wave run up of 119.7 feet PSD at the power block and 126.1
feet PSD at the SWIS.3 The LAR also states that the PMHSS can be screened from the RICT program using the initial screening criterion C1, Plant damage potential is less than events for which the plant is designed, following the closure of all of the watertight doors listed in attachment 1 of procedure SC.OP-AB.ZZ-0001. Section 5.5.3 of enclosure 4 of the TSTF-505 LAR states that the local intense precipitation (LIP) flood causing mechanism was found not to be bounded by the CDB following the completion of the FHRR, so a focused evaluation (FE) was performed to determine if Salem has adequate flood protection from the LIP mechanism, which concluded that Salems flood protection from the PMHSS bounds the protection required for the LIP. The LAR also states that the margin between the two hazards is at least 10 feet of elevation (102.0 feet PSD per LAR enclosure 4, table E4-9 versus 120.4 feet PSD at the power block and 127.3 feet PSD at the SWIS building) and that, with all the doors in attachment 1 of procedure SC.OP-AB.ZZ-0001 closed and clear procedural guidance and warning time to complete, LIP can be screened from the RICT program using criterion C1, as confirmed in the FE. Procedure SC.OP-AB.ZZ-0001 states in a note that doors, hatches or manway covers which are taken credit for in the Updated Final Safety Analysis Report (UFSAR) as watertight external flood barriers below the design basis flood height of 120.4 feet are identified with Technical Specification (T/S) 3.7.5.1/3.7.5 in Attachment 1, T/S Protective Doors, and that these doors, hatches, or manway covers should be given priority for closure within 2 hours per T/S 3.7.5.1/3.7.5. In light of these observations, address the following: a) Are the doors and penetrations credited in the UFSAR as watertight external flood barriers below the design basis flood height considered to be watertight in their default configuration or do operators need to perform actions (other than verifying they are closed) for the doors and penetrations considered to be watertight? b) Are the doors and penetrations credited in the UFSAR as watertight external flood barriers below the design basis flood height normally closed? If so, describe how often they are confirmed to be closed (e.g., operator rounds, surveillance tests, etc.). c) The following questions relate to the doors that are credited to be closed in two hours in attachment 1 of procedure SC. OP-AB.ZZ-0001. (i) How long can it take for the floodwater to reach plant grade? (ii) How many operators are needed to close the doors and penetrations? (iii) Are these operators always onsite at the time of the event or do they need to be called in to perform these actions? Audit Question-32 (APLC-03) On page 8 of the 10 CFR 50.69 LAR, Table 3-1 Categorization Evaluation Summary, lists the seismic hazard twice in the Risk (non-modeled) section. The first time it is listed as part of other external hazards, and the second time it is listed individually. 3 Public Service Electric and Gas Company datum (PSD) +89 feet is equivalent to 0 feet mean sea level.
If this is a typographical error, provide an updated table with an appropriate correction. If this is not a typographical error, explain if seismic risk will be evaluated at both the function and component level in a way that the Integrated Decision-making Panel (IDP) allows for changes from HSS to LSS. Audit Question-33 (APLC-04) NEI 00-04, Revision 0, Section 5.4, Assessment of Other External Hazards, provides guidance on assessment of other external hazards (excluding fire and seismic) in 10 CFR 50.69 categorization of SSCs. Specifically, Figure 5-6, Other External Hazards, of NEI 00-04 illustrates a process that begins with an SSC selected for categorization and proceeds through a flowchart for each external hazard. Figure 5-6 indicates that if a component participates in a screened scenario, for that component to be considered a low safety significant item, it has to be further shown that the screened scenario would not become unscreened if the component were removed. Section 3.2.4, Other External Hazards, of the enclosure, Evaluation of the Proposed Change, to the LAR indicates that all other external hazards besides seismic events were screened with, Other External Hazards Disposition, of the enclosure to the LAR providing the results. Based on this description, it appears to the NRC staff that when an SSC is categorized it will not be evaluated using the guidance in NEI 00-04, Figure 5-6, to confirm that the SSC is not credited in screening an external hazard. The NRC staff notes that plant changes, operating experience, errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard. Therefore, address the following: a) Clarify whether an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6, to confirm that the SSC is not credited in screening an external hazard. b) If an SSC will not be evaluated using the guidance in NEI 00-04, Figure 5-6, to confirm that the SSC is not credited in screening an external hazard at the time of categorization, explain how plant changes, operating experience, and errors or limitations in the PRA models that could change that decision are addressed. Audit Question-34 (APLC-05) Section 2.3.1, item 7, of NEI 06-09, Configuration Risk Management Process and Application of Technical Specifications, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs safety evaluation for NEI 06-09 states that where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT. In Enclosure 4 of the LAR, Section 3, Seismic Risk Contribution Analysis, the licensee provides its seismic core damage frequency (SCDF) penalty value of 2.3E-6/year for use in this application based on the plant-specific seismic hazard curves submitted by the licensee in response to the NRCs post-Fukushima actions. However, the licensee did not provide information on the plant-level high confidence of low probability of failure (HCLPF) value used for the SCDF estimate associated with the composite beta factor (c) of 0.84 or the de-inerted containment SLERF value in the LAR. The NRC staff noted that the licensees PSEG calculation
SA-MISC-032, Seismic CDF and LERF Penalty Estimates for the Salem TSTF-505 (RICT) Program, in the audit portal, provides information on the HCLPF and c values used and how they are determined and used in estimating the SCDF, but it does not specifically give the HCLPF value. Provide on the docket a summary of the HCLPF values, c values, and de-inerted containment SLERF value used for SCDF and LERF penalty estimates for this application with justification that these values are the best-available representation of the plants seismic capacity. Audit Question-35 (EICB-01) Justify under condition 21, TS Table 3.3-3 Function 8.f is not a loss of function. Audit Question-36 (EICB-02) of the Salem TSTF-505 LAR states, The description of proposed changes to the protective instrumentation and control features in TS Section 3.3, "Instrumentation," should confirm that at least one redundant or diverse means (other automatic features or manual action) to accomplish the safety functions (for example, reactor trip, SI, containment isolation, etc.) remains available during use of the RICT, consistent with the defense-in-depth philosophy as specified in RG 1.174. (Note that for each application, the staff may selectively audit the licensing basis of the most risk significant functions with proposed RICTs to verify that such diverse means exist.) Sections 2.0 and 3.0 show that for each reactor trip system (RTS) function and each engineered safety features actuation system (ESFAS) instrument function in TS 3/4.3, Instrumentation, there is at least one diverse means for initiating the safety function. Provide a diversity justification that describes other means that exist to initiate the safety function for each plant accident condition that each affected I&C function is currently designed to address. The evaluation of diverse means, should identify the conditions that the functional unit responds to, and for each condition, other means (e.g., diversity, redundancy, or operator actions) that can be used. Technical Review Branch/Review Area APLA - PRA Licensing Branch A APLA - PRA Licensing Branch B APLA - PRA Licensing Branch C EICB - Instrumentation and Controls Branch
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