ML23335A065

NRR E-mail Capture - GGNS - Audit Questions for LARs Associated with TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Init. 4b and 50.69, Risk-Informed Cat. and Treatment of SSCs for Nuclear Pwr Rx (L-2023-LLA-0081/0080)

ML23335A065
Person / Time
Site:	Grand Gulf
Issue date:	12/01/2023
From:	Siva Lingam NRC/NRR/DORL/LPL4
To:	Meister R Entergy Nuclear Operations
References
L-2023-LLA-0081, L-2023-LLA-0080
Download: ML23335A065 (28)

Text

From:

Siva Lingam Sent:

Friday, December 1, 2023 10:05 AM To:

Meister, Richard Cc:

Jennifer Dixon-Herrity; Jennie Rankin; Ching Ng; Hardy, Jeffery A

Subject:

GGNS - Audit Questions for LARs Associated with TSTF-505, Provide Risk-Informed Extended Completion Times-RITSTF Initiative 4b and 10 CFR 50.69, Risk-Informed Categorization and Treatment of SSCs for Nuclear Power Reactors (L-2023-LLA-0081/0080)

Attachments:

Audit Questions (TSTF-505 and 50.69).docx Attached please find the official audit questions for the subject license amendments.

Siva P. Lingam U.S. Nuclear Regulatory Commission Project Manager Palo Verde Nuclear Generating Station, Units 1, 2, and 3 Grand Gulf Nuclear Station Entergy Fleet Location: O-9E22; Mail Stop: O-9E03 Telephone: 301-415-1564 E-mail address: Siva.Lingam@nrc.gov

Hearing Identifier:

NRR_DRMA Email Number:

2323 Mail Envelope Properties (SJ0PR09MB6109B03BDB0887D2A53666BAF681A)

Subject:

GGNS - Audit Questions for LARs Associated with TSTF-505, Provide Risk-Informed Extended Completion Times-RITSTF Initiative 4b and 10 CFR 50.69, Risk-Informed Categorization and Treatment of SSCs for Nuclear Power Reactors (L-2023-LLA-00810080)

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12/1/2023 10:04:00 AM From:

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Siva.Lingam@nrc.gov Recipients:

"Jennifer Dixon-Herrity" <Jennifer.Dixon-Herrity@nrc.gov>

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Tracking Status: None "Ching Ng" <Ching.Ng@nrc.gov>

Tracking Status: None "Hardy, Jeffery A" <jhardy@entergy.com>

Tracking Status: None "Meister, Richard" <rmeist1@entergy.com>

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1 QUESTIONS FOR THE AUDIT RELATED TO THE LICENSE AMENDMENT REQUESTS TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF505, REVISION 2, AND IMPLEMENT 10 CFR 50.69 ENTERGY OPERATIONS, INC GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 By letters dated June 6, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML23158A043 and ML23158A044, respectively), Entergy Operations, Inc. (Entergy, the licensee) submitted two license amendment requests (LARs) for Grand Gulf Nuclear Station, Unit 1 (Grand Gulf, GGNS). The proposed amendments would modify Renewed License No. NPF-29 and the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times, RITSTF [Risk-Informed Technical Specification Task Force]

Initiative 4b (ML18183A493), and to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), section 50.69, Risk-informed categorization and treatment of structures, systems, and components [SSCs] for nuclear power reactors. On August 18, 2023, the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan (ML23226A100) that conveyed intent to conduct a regulatory audit during the week of December 4, 2023, to support its review of the above license amendments. 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 will address both LARs.

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ML17317A256), states that the scope, level of detail, and technical adequacy of the probabilistic risk assessment (PRA) are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process. The NRCs safety evaluation (SE) for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, dated November 6, 2006 (ML122860402),

and the NRCs Final SE for NEI 06-09-A, dated May 17, 2007 (ML071200238), state that the PRA models should conform to the guidance in RG 1.200, Revision 1, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. The version applicable to this LAR is RG 1.200, Revision 2 (ML090410014), which states that the quality of the PRA must be compatible with the safety implications of the proposed change and the role the PRA plays in justifying the change.

The provisions of 10 CFR 50.69 allow adjustment of the SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design basis functions. In accordance with 10 CFR 50.69, a risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four Risk-Informed Safety Classifications (RISC) categories. The SSCs are classified as having either high-safety-significant (HSS) functions (i.e., RISC-1 and RISC-2 categories) or low-safety-significant (LSS) functions (i.e.,

2 RISC-3 and RISC-4 categories). For HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment. For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d).

NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ML052910035), as endorsed by RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (ML061090627) provides an acceptable process for determining the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

The NRC staff has determined that additional information is needed to support its review as shown in the following audit questions. The questions are identified by technical review branch/area followed by number in sequence. The common audit questions for both LARs are listed first, followed by TSTF-505 LAR and 10 CFR 50.69 LAR.

Common Audit Questions for Both LARs (TSTF-505 and 10 CFR 50.69)

APLA PRA Credit for FLEX Equipment and Actions NRC memorandum dated May 6, 2022,1 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.

With regards to equipment failure probability, in the memorandum dated May 6, 2022, the NRC staff states in Conclusion 4:

Licensees that choose not to use the generic failure probabilities in [Pressurized Water Reactor Owners Group] PWROG-10842 [FLEX Equipment Data Collection and Analysis]

to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRA used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval.

With regards to the uncertainty related to equipment failure probabilities, in the updated NRC memorandum, the NRC staff states in Conclusion 8:

PWROG-18043, Revision 1, [FLEX Equipment Data Collection and Analysis] notes that there was insufficient data to quantify the failure to load probabilities for portable diesel generators due to lack of detailed data. To account for the uncertainty in the testing activitieslicensees should ensure their preventive maintenance strategies include such testing and that the data reported provides this information. licensees should continue to assess the uncertainty in equipment failure rates and address or disposition it.

With regards to human reliability analysis (HRA), in the memorandum dated May 6, 2022, the NRC staff states, in part, in Conclusion 11: ML052910035 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).

3 EPRI [Electric Power Research Institute] 3002013018 [Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Capability (FLEX) and Use of Portable Equipment: Examples and Guidance] provides updated detailed industry guidance for estimating the human error probabilities (HEPs) of the actions needed to implement mitigating strategies using portable equipment.

EPRI 3002013018 provides guidance that is acceptable to the NRC, with the clarifications provided in the May 6, 2022, memorandum.

With regards to PRA Upgrade, the staff states in the updated memorandum in Conclusion 2:

Therefore, Conclusion 2 remains unchanged [that] for any new risk-informed application that has incorporated mitigating strategiesthe licensee should either perform a focused-scope peer review of the PRA model or demonstrate [that it does not meet the three criteria of an PRA Upgrade].

, section 2 of the TSTF-505 LAR states that the fire PRA takes very minimal credit for FLEX strategies, which includes reactor core isolation cooling (RCIC) injection from the upper containment pool and powering the H2 igniters. It is unclear to the NRC staff what the phrase very minimal credit for FLEX strategies means and whether the internal events (including internal flooding) PRA also credits any FLEX equipment and actions. Section 3.0 of the 50.69 LAR states that the PRA models described in the 50.69 LAR are the same as those described within the TSTF-505 LAR. The first disposition in Attachment 6 of the 50.69 LAR states, the PRA models do not credit mitigating FLEX strategies. However, section 2.0 of enclosure 9 of the TSTF-505 LAR states, the fire PRA model takes very minimal credit for FLEX strategies. Furthermore, Entergy Report PSA-GGNA-08-UNR, GGNS Impact of Model Uncertainty to RICT [risk informed completion time] Process, highlights FLEX as an area of uncertainty. It is unclear to the NRC staff whether the PRA models used for supporting the 50.69 program (i.e., internal events, internal flooding, internal fire) credit FLEX strategies, and if so, to what extent they are credited. Also, it appears from portal reports that two portable diesel generators and two portable diesel driven pumps are credited in the PRA models.

For the Grand Gulf PRA models, address the following:

a) Describe the FLEX equipment and actions credited in the Grand Gulf PRA models. If there is a difference between the internal events and fire PRA models, then explain that difference.

b) Describe the approach used to determine the failure probabilities of credited FLEX equipment. The discussion should include:

i.

Justification that the parameter values for portable equipment are in alignment with the guidance in the NRC memorandum dated May 6, 2022.

ii.

Provide an assessment, such as a sensitivity study, of the impact of the uncertainty in FLEX equipment failure rates credited in Grand Gulf PRA models on the 10 CFR 50.69 SSC categorization results. Include in this discussion, the impact of FLEX on exceeding importance rankings for affected SSCs and whether the uncertainty associated with FLEX modeling is a key source of uncertainty for 10 CFR 50.69. If this

4 uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed for 10 CFR 50.69 categorization (e.g.,

identify specific PRA sensitivity studies that will be performed for SSC categorization and the results of these sensitivity studies will be given to the Integrated Decision-making Panel (IDP) for consideration in the final risk characterization of SSCs).

This assessment should include, if applicable, any modifications to FLEX modeling performed to address the NRC clarification in the May 6, 2022, memorandum.

iii.

Provide an assessment, such as a sensitivity study, of the impact of the uncertainty in FLEX equipment failure rates credited in Grand Gulf PRA models on the RICT program.

Entergy Report PSA-GGNS-01-HR-01, Probabilistic Risk Assessment Human Reliability Analysis, appears to indicate that if FLEX is not credited, the core damage frequency (CDF) would increase by 8.3 percent and the large early release frequency (LERF) by 16.4 percent. The NRC staff notes that the impact on certain RICT configurations could be even greater.

This assessment should include, if applicable, any modifications to FLEX modeling performed to address the NRC clarification in the May 6, 2022, memorandum.

c) Describe the approach used to model operator actions associated with implementing FLEX equipment and the plant staff that perform these actions.

The discussion should include:

i. Discussion of how the licensee evaluated the impact of the NRC clarifications cited above compared to EPRI 3002013018 on FLEX HRA methodology.

ii. Provide updated FLEX HRA results, if applicable, to address the NRC clarifications in the May 6, 2022, memorandum.

d) Provide justification that the modeling of FLEX equipment and actions do not meet the definition of a PRA Upgrade as defined by RG 1.200.

-OR-Propose a mechanism that commits to conduct a focused-scope peer review (FSPR) of (FSPR) of (FSPR) of the modeling of FLEX equipment and actions in the Grand Gulf PRA models. Include in the mechanism a commitment to close out all finding-level facts and observations (F&Os) that may result from the FSPR prior to implementing the RICT process.

APLA Open Phase Condition

5 Section C.1.4 of RG 1.200 states the base (e.g., Model of Record) PRA is to represent the as-built, as-operated plant to the extent needed to support the application. The licensee is to have a process that identifies updated plant information that necessitates changes to the base PRA model.

In response to the January 30, 2012, Open Phase Condition (OPC) event at the Byron Station, the NRC issued Bulletin 2012-013. As part of the initial Voluntary Industry Initiative (VII) for mitigation of the potential for the occurrence of an OPC in electrical switchyards4, licensees have made the addition of an Open Phase Isolation System (OPIS). In accordance with Staff Requirements Memorandum (SRM)-SECY-16-00685, the NRC staff was directed to ensure that licensees have appropriately implemented OPIS, and that licensing bases have been updated accordingly. From the revised voluntary initiative6 and the resulting industry guidance in NEI 19-027 on estimating the risk associated with an OPC and OPIS risk, it is understood that the risk impact of an OPC can vary widely, depending on electrical switchyard configuration and design. In light of this observation, provide the following information:

a) A discussion of the impact on risk of the OPC issue at Grand Gulf, Unit 1.

b) Discuss whether modeling of the OPC issue and any OPIS that has been installed and implemented at Grand Gulf have been, or are planned to be, incorporated as part of the plant model of record (MOR). If so, provide the following:

i. The schedule for the inclusion of OPC and OPIS modeling in the MOR.

ii. The impact, if any, to key assumptions and sources of uncertainty.

iii. A discussion of the HRA methods and assumptions used for OPIS alarm manual response.

iv. The impact to external events (e.g., fire, seismic, flooding, high winds, tornado, other external events, etc.).

v. A discussion of the risk impact of inadvertent OPIS actuation and justification for its exclusion.

c) If OPC and OPIS are not planned to be included in the MOR, provide justification why the risk impact is not included by performing either a qualitative or sensitivity analysis.

APLA/APLB Deviations from NRC Endorsed Guidance as Source of Modeling Uncertainty 3 U.S. NRC Bulletin 2012-01, Design Vulnerability in Electric Power System (ML12074A115).

4 Anthony R. Pietrangelo to Mark A. Satorius, Ltr re: Industry Initiative on Open Phase Condition -

Functioning of Important-to-Safety Structures, Systems and Components (SSCs), dated October 9, 2013 (ML13333A147).

5 U.S. NRC SRM-SECY-16-0068, Interim Enforcement Policy for Open Phase Conditions in Electric Power Systems for Operating Reactors, dated March 9, 2017 (ML17068A297).

6 Doug True to Ho Nieh, Ltr re: Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ML19163A176).

7 Nuclear Energy Institute (NEI) 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, Revision 0, April 2019 (ML19122A321).

6 RG 1.200 states NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. The relatively extensive and detailed reviews of fire PRAs undertaken in support of LARs to transition to NFPA-805 determined that implementation of some of the complex fire PRA methods often used non-conservative and over-simplified assumptions to apply the method to specific plant configurations. Some of these issues were not always identified in F&Os by the peer review teams but are considered potential key assumptions by the NRC staff because using more defensible and less simplified assumptions could substantively affect the fire risk and fire risk profile of the plant. The NRC staff evaluates the acceptability of the PRA for each new risk-informed application and as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision-making may vary with the relative weight given to the risk assessment element of the decision-making process. The NRC staff notes that the calculated results of the PRA are used directly to calculate a RICT, which subsequently determines how long SSCs (both individual SSCs and multiple, unrelated SSCs) controlled by Technical Specifications can remain inoperable. Also, the calculated results of the PRA are used in the 10 CFR 50.69 SSC categorization process. Therefore, the PRA results are given a high weight in a TSTF-505 and 10 CFR 50.69 application, and the NRC staff requests additional information on the following issues that have been previously identified as potentially key fire PRA assumptions.

Use of Unacceptable Methods a) The LARs provide the history of the fire PRA peer review but does not discuss methods used in the fire PRA. Methods may have been used in the fire PRA that deviate from guidance in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities" (ML052580075, ML052580118, and ML103090242), or other acceptable guidance (e.g., frequently asked questions (FAQs), NUREGs, or interim guidance documents).

i.

Identify methods used in the fire PRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance.

ii.

If such deviations exist, then justify their use in the fire PRA and impact on the RICT and SSC categorization results.

OR As an alternative, add an implementation item(s) to replace those methods with a method acceptable to NRC prior to the implementation of the RICT Program and the 10 CFR 50.69 program. Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

Reduced Transient Heat Release Rates (HRRs) b) The key factors used to justify using transient fire reduced heat release rates (HRRs) under those prescribed in NUREG/CR-6850 are discussed in the .F. Assessment Plan Available for Review|June 21, 2012, letter]] from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, "Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires" (ML12171A583). Entergy report PSA-GGNS-03-FS, Grand Gulf Nuclear Station Fire PRA Scoping Modeling Report, appears to indicate that the fire PRA uses the newer HRR distribution from NUREG-2178, Refining

7 And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) (EPRI 3002005578) (ML15111A045), and in other cases HRRs from NUREG/CR-6850 were used. Accordingly, it is not clear whether reduced NUREG/CR-6850 HRRs were used. Given these observations, address the following:

If any reduced transient HRRs below the bounding 98 percent HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs.

Include in this discussion:

i.

Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

ii.

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.

iii.

The results of a review of records related to compliance with the transient combustible and hot work controls.

Treatment of Sensitive Electronics c) 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. Entergy report PSA-GGNS-03-FS states that for modeling failure of sensitive electronics due to heat flux the guidance from FAQ 13-0004 was used unless the sensitive electronic components did not meet exclusion guidelines. The report states that in this case, the criteria from NUREG/CR-6850, Appendix H.2 was used (a radiant heat flux of 3kW/m2 or a temperature of 65°C.). It is not completely clear what the phrase exclusion guidelines meant. Therefore, address the following:

i.

Confirm that phrase exclusion guidelines refers to the caveats in FAQ 13-0004, about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents) and confirm through walkdowns and other information how the configuration of cabinets containing sensitive electronics was established.

ii.

If the approach used by Grand Gulf cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no consequential impact on the RICT calculations and SSC categorization results.

OR iii.

As an alternative, add an implementation item(s) to replace the current approach with an acceptable approach prior to the implementation of the RICT Program and the 10 CFR 50.69 program. Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

8 Obstructed plume model d) NUREG-2178, Volume 1, contains refined peak HRRs, compared to those presented in NUREG/CR-6850 and guidance on modeling the effect of plume obstruction. However, NUREG-2178 stipulates 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.

Entergy report PSA-GGNS-03-FS indicates that obstructed plume modeling was used but does not describe the elevation assumed for the base of the plume. Therefore, address the following:

i.

Where the obstructed plume modeling was used, indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.

ii.

Justify any modeling in which the base of an obstructed plume is located at less than one-half of the cabinet's height.

OR As an alternative to part (ii) above, remove the credit or add an implementation item(s) to remove credit for the obstructed plume model in the fire PRA prior to the implementation of the RICT program and the 10 CFR 50.69 program.

Systems not credited in the fire PRA e) The NRC SE of NEI 06-09, Revision 0-A states:

When key assumptions introduce a source of uncertainty to the risk calculations (identified in accordance with the requirements of the ASME standard), TR NEI 06-09, Revision 0, requires analysis of the assumptions and accounting for their impact to the RMTS calculated RICTs.

The NRC SE of NEI 00-04 (RG 1.201, ML061090627) states the purpose of the sensitivity studies as part of the risk categorization process in the 10 CFR 50.69 program is to address the impact of parameter and model uncertainties on the categorization.

Table 2, Item No. 1, of Entergy reports, PSA-GGNS-08-UNR, Revision 0, and PSA-GGNS-08-UNC, Revision 0, GGNS Impact of Model Uncertainty to 10 CFR 50.69 Categorization Process, explains certain components were assumed to always fails.

The reports state that basic events for which the component cable routing are not known, developed, or assumed are modeled to fail for every scenario (titled Ever Failed events) except for Main Control Room abandonment scenarios. The report PSA-GGNS-08-UNR states that this conservatism would tend to result in more conservative RICT times, and was therefore screened.

Though the assumption used in the base fire PRA model about failing components having no cable routing information is conservative, NRC staff notes that this conservatism in PRA modeling could have a nonconservative impact on the RICT calculations. If an SSC (in the RICT program) is part of a system not credited in the fire PRA or supports a system that is subject to a RICT, then the increase in risk due to

9 taking that SSC out of service is masked. Also, conservatism in PRA modeling could mask SSC categorizations in the 10 CFR 50.69 program. Therefore, address the following:

i.

Identify the systems or components that are assumed to be always failed in the fire PRA, or are not included in the fire PRA, due to lack of cable tracing or other reasons.

ii.

Justify that this assumption has an inconsequential impact on the RICT calculation and SSC categorization results.

iii.

If in response to part (ii) above, it cannot be determined that the cited assumption has an inconsequential impact on the estimated RICTs, then identify what programmatic changes will be considered to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional Risk Management Actions (RMAs)).

iv.

If in response to part (ii) above, it cannot be determined that the cited assumption has an inconsequential impact on the SSC categorization results, then describe and provide a basis for how this uncertainty will be addressed for SSC categorization under the 10 CFR 50.69 program (e.g., identify specific PRA sensitivity studies that will be performed for SSC categorization and the results of these sensitivity studies will be given to the IDP for consideration in the final risk characterization of SSCs).

Well-Sealed Motor Control Center cabinets f) Guidance in FAQ [Frequently Asked Question] 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440 V. 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. Section 6 of NUREG/CR-6850 states that well-sealed cabinets below 440V should be excluded from the Bin 15 count. Counting these cabinets has the effect of diluting the count and decreasing the ignition frequency. Treatment of well-sealed cabinets less than 440V does not appear to be addressed in Entergy report PSA-03-FS, Grand Gulf Nuclear Station Fire PRA Scoping Modeling Report or PSA-GGNS-03-IGN, Grand Gulf Nuclear Station Fire PR Fire Ignition Frequency. Therefore, address the following:

If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, provide justification for using this approach as this is contrary to the guidance.

Audit Questions for TSTF-505 LAR APLA Accounting for the State of Knowledge Correlation (SOKC)

Based on RG 1.174 and section 6.4 of NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Main Report," dated March 2017 (ML17062A466), for a Capability Category II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance

10 guidelines. The risk management threshold values for the RICT program have been developed based on RG 1.174, and therefore, the most appropriate measures with which to make a comparison are also mean values. Point estimate PRA results are commonly calculated and reported, but these are typically lower than the mean values and do not account for the state-of-knowledge correlation (SOKC) between nominally independent basic event probabilities. Failure probabilities that are derived from common data sources should be correlated in a parametric uncertainty analysis. NUREG-1855, Revision 1, provides guidance on evaluating how the SOKC uncertainty impacts the comparison of the PRA results with the guideline values. 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 or the difference between point estimate and mean would have a minimal impact on the application).

of the LAR presents the Grand Gulf total CDF and LERF as well as the internal events, internal flooding, fire, and seismic penalty CDF and LERF contributors. The LAR does not indicate whether these values are point estimates or if the mean values determined from parametric uncertainty analysis accounts for the SOKC. The total risk results are well below the RG 1.174 risk guidelines of 1E-04 per CDF and 1E-05 per year LERF, so the use of point estimate does not impact whether the total CDF and LERF risk acceptance guidelines values might be exceeded.

However, the LAR does not discuss how the SOKC correlation is addressed to meet the ASME/ANS PRA standard Supporting Requirement (SR) QU-A3 or explicitly state whether the RICT program will be administered using point estimate values or mean values.

Therefore, if the RICT program will be administered using point estimate values rather than mean values determined from parametric uncertainty analysis, then address the following to demonstrate that impact of using point estimate values has an inconsequential impact on the RICT calculations:

a. Discuss and provide the results of a comparison study between the RICT values calculated using point estimate versus mean risk values for various Limiting Conditions for Operation (LCOs) conditions in the scope of the RICT program. The LCO conditions selected for this comparison study should be those judged most likely to be impacted by the SOKC uncertainty and have a point estimate RICT less than the 30-day backstop so that the comparison results are not masked by the backstop. Provide the bases for the chosen LCO conditions in this comparison study.

b. For the fire PRA clarify how SOKC was addressed by identifying the fire parameters that were correlated in the parametric uncertainty analysis.

APLA PRA Model Uncertainty Analysis Process The NRC staff SE to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and sources of uncertainty and to assess and disposition each as to their impact on the RMTS application. NUREG-1855, Revision 1, dated March 2017 (ML17062A466) presents guidance on the process of identifying, characterizing, and qualitative screening of model uncertainties.

LAR enclosure 9, section 1 indicates that the process used to evaluate sources of uncertainty for the RICT application follows the guidance illustrated in Figure 4-1 of EPRI TR-1016737,

11 Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments. The LAR states that for the internal events PRA both plant specific sources of uncertainty from the PRA notebooks and generic sources of uncertainty from EPRI TR-1016737 were considered.

Guidance from NUREG-1855, Revision 1 and EPRI TR-1026511, "Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty" is not cited in the LAR (though portal documents uncertainty analysis cite it. EPRI TR-1026511 provides generic sources of uncertainty for both the fire PRA and Level 2 PRA. Also, the LAR does not explain whether both plant specific and generic sources uncertainty were used to identify key sources of uncertainty. Finally, Criterion No. 4 of section 2 of the LAR states that sensitivity studies were performed to confirm that the impact of the source of uncertainty had a negligible impact on the RICT program, but no sensitivity study results are reported in the LAR. Accordingly, details of Entergys process for identifying a comprehensive list of sources of uncertainty and then evaluating key sources of uncertainty are not clear to NRC staff. Given these observations, address the following:

a) Clarify how Entergys process for identifying key sources PRA model uncertainty aligns with the guidance in NUREG 1855 Revision 1.

b) Clarify how a comprehensive listing of fire PRA model and Level 2 PRA model sources of uncertainty were compiled for evaluation.

i.

Include explanation of whether the listing included consideration of both plant-specific sources and generic sources of uncertainty from EPRI TR-1026511.

ii.

If a comprehensive listing of fire PRA and Level 2 PRA source of model uncertainty does not include evaluation of plant specific and generic sources of uncertainty, then provide an assessment of plant specific and generic sources of uncertainty demonstrating that they have an inconsequential impact or the RICT calculations.

c) Beyond the sensitivity study discussed in section 2 of enclosure 9 of the LAR on the loss of decay heat removal contribution to LERF, describe any other sensitivity studies performed to evaluate key sources of model uncertainty and summarize their results.

Include justification that the results of the sensitivity studies demonstrate that the candidate key sources of uncertainty have an inconsequential impact on the RICT calculations.

APLA/APLB Dispositions of PRA Model Assumptions and Sources of Uncertainty The NRC staff SE to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and sources of uncertainty and to assess and disposition each as to their impact on the RMTS application. NUREG-1855, Revision 1, presents guidance on the process of identifying, characterizing, and qualitative screening of model uncertainties.

LAR enclosure 9 Table E9-1 provides dispositions for only three candidate key assumptions and sources of uncertainty, and all were associated with translation of the baseline PRA models to the Configuration Risk Management Program (CRMP) model. Enclosure 9 explains that except for one case there were no sources of internal events or fire PRA model uncertainty that required a sensitivity study to determine its impact on the RICT calculations. However, NRC staff reviewed uncertainty analysis portal documents including PSA-GGNS-08-UNR, Revision 0,

12 GGNS Impact of Model Uncertainty to RICT Process and identified sources of uncertainty for which further information is needed to understand their disposition for this application, including whether a RMA may be required. Therefore, address the following:

a) Entergy report PSA-GGNS-08-UNR, Table 1, Item 5 states that the current modeling associated with HPCS and RCIC is conservative. The report states the HRA dependencies between HPCS and RCIC actions and the action to manually depressurize are conservative because these HRA combinations do not recognize the intervening success for inhibiting the Automatic Depressurization signals. Therefore, the uncertainty was evaluated as not being identified as a key source of uncertainty. The NRC staff observes that conservatism in PRA modeling could possibly have a nonconservative impact on the RICT calculations particularly since HPCS and RCIC are in the RICT program. In general, if the reliability of systems or components in the RICT program are not credited or fully credited in the PRA then the full risk increase from taking that system or component out of service could be masked. Therefore, address the following:

i.

Justify that the cited assumption has an inconsequential impact on the RICT calculation. If the justification cannot be made qualitatively, then provide a quantitative assessment such as a sensitivity study demonstrating that the impact of this modeling on affected RICTs is inconsequential.

ii.

If, in response to part (i) above, it cannot be determined that the cited assumption has an inconsequential impact on the estimated RICTs, then identify if programmatic changes will be considered to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional RMAs).

b) Entergy report PSA-GGNS-08-UNR, Table 2 Item 5 states that crediting fire wrap inappropriately could result in non-conservative risk results. Entergy report PSA-GGNS-03-FQ, Grand Gulf Nuclear Station Fire PRA Quantitation, section 5.6.3 states that fire wrap is important in Division I and Division II switchgear rooms to reduce the likelihood of having a Station Blackout due to Failures of ESF11 and ESF-21. Entergy report PSA-GGNS-03-FQ dispositions this source of uncertainty by stating the fire wrap credit is in-alignment with current industry accepted approaches.

Given the stated importance of this source of uncertainty, describe the fire wrap and identify the fire protection codes to which the fire wrap conforms for the credit taken in the switchgear rooms for the fire PRA. If the conformance involves demonstration of equivalency rather than straightforward compliance, then discuss the corresponding engineering assessment.

APLA Digital Instrumentation and Control Modeling NEI 06-09 states the following concerning the quality of the PRA model 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 instrumentation and control (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

13 industry data for digital I&C components, the difference 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 whether the licensee credited digital system in the PRA models that will be used in the RICT program or whether this modeling can impact the RICT calculations. Therefore, address the following:

a) Identify any digital I&C systems 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 has an inconsequential impact on the RICT calculations.

OR Identify if programmatic changes will be considered to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional RMAs).

APLA Impact of Seasonal Variations The Tier 3 requirement of RG 1.177 stipulates that a licensee should develop a program that ensures that the risk impact of out of service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09 and the NRC SE to this guidance state that for the impact of seasonal changes either conservative assumptions should be made or, the PRA should be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration.

LAR enclosure 8 on attributes of the Real Time Risk (RTR) model, section 2 states:

The impact of outside air temperatures on system requirements is addressed in the CRMP model. There are no changes in the PRA success criteria based on the time in the core operating cycle.

The CRMP model accounts for severe weather conditions.

These statements seem to indicate that modeling adjustment (e.g., in success criteria) are made to account for seasonal variation. However, it is not clear to NRC staff how adjustments will be made in the RTR model to account for impact of seasonal variation, or what criteria will be used to know when PRA adjustments are needed. Therefore, address the following:

a) Discuss the modeling that will be subject to adjustment due to seasonal variations such as hot or cold weather or other environmental factors (e.g., water levels) that can impact the performance of plant systems. Explain what kind of adjustments will be made and clarify whether they will be made conservatively like the adjustments that might be made for HVAC dependency.

b) Explain what criteria (e.g., triggers) are used to know when PRA adjustments due to seasonal variations need to be made in the RTR and justify that this approach is consistent with the guidance in NEI 06-09.

14 APLA In-Scope LCOs and Corresponding PRA Modeling The NRC SE to NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. Enclosure 1, Table E1-1 of the LAR identifies each TS LCO proposed for the RICT program, describes whether the systems and components participating in the TS LCO are implicitly or explicitly modeled in the PRA, and compares the design basis and PRA success criteria. For certain TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some LCO conditions, the LAR did not provide enough description for NRC staff to conclude that the PRA modeling will be sufficient for each proposed LCO Condition. Address the following:

a) LAR Table E1-1 indicates for TS LCO 3.3.1.1.A.1 regarding Intermediate Range Monitors, TS LCO 3.3.1.1.A.2 regarding Average Power Range Monitors, TS LCO 3.3.1.1.A.7 regarding Drywell Pressure - High, and TS LCO 3.3.1.1.A.11 regarding Reactor Mode Switch - Shutdown Position that no modeling exists in the PRA for the channels associated with these instruments. The LAR states for each of these LCOs conservative surrogate modeling will use both RPS mechanical seal-in relays of the affected subsystem (e.g., K14A and K14E). It appears that if more than one of these LCOs enters a RICT condition at the same time, then the cited surrogate modeling is insufficient to account for the total risk increase because the same surrogate is used.

Given these observations, address the following:

i.

Explain how the surrogate modeling would be performed that accounts for the risk increase in Completion Time extensions if more than one RICT from the set cited above is entered during the same time interval.

ii.

If multiple RICTs from the set of cited LCOs would not be allowed to be entered at overlapping times, then explain the mechanism in the RICT program that prevents this from occurring.

b) LAR Table E1-1 states in the seventh column for TS LCO 3.3.1.1.B.1 regarding Reactor Protection System instrumentation functions that the PRA success criteria is the same as the design basis success criteria. However, in the eighth column the LAR defines conservative surrogate modeling. Reconcile this apparent inconsistency.

c) LAR Table E1-1 states in the seventh column for TS LCO 3.3.5.1.F regarding Automatic Depressurization System (ADS) initiation logic and instrumentation Functions 4.d and 5.d for Reactor Vessel Water Level - Low Level 3 that the PRA success criteria is the same as the design basis success criteria. However, in the eighth column the LAR defines a conservative surrogate. Reconcile this apparent inconsistency.

d) LAR Table E1-1 states in the seventh column for TS LCO 3.3.6.1.A.2 regarding Primary Containment and Drywell Isolation that the PRA success criteria is the same as the design basis success criteria. However, in the last column the LAR defines conservative surrogate modeling. Reconcile this apparent inconsistency.

e) LAR Table E1-1 states in the seventh column for TS LCO 3.6.1.3.A regarding Primary Containment Isolation Valves that the PRA success criteria is the same as the design

15 basis success criteria. However, in the eighth column the LAR defines conservative surrogate modeling. Reconcile this apparent inconsistency.

f) LAR Table E1-1 indicates for TS LCO 3.6.1.6.A regarding the Low Low Set (LLS) valves that the valves are not modeled, and therefore a conservative surrogate will be used.

The LAR states for this LCO:

Affected SRV [Safety Relief Valve] will be failed and SRV failure to reclose probability will be increased as a conservative surrogate (see Note 8)

Note 8 to Table E1.1 referred to above states:

LCOs impacting LLS function of SRV will be treated as the affected SRV is inoperable for all functions AND the SRV Fail to Reclose failure mode probability will be doubled to account for additional SRV cycles. This is conservative, because one LLS valve can be inoperable without impact.

Similarly, for LAR Table E1-1 presents for TS LCO 3.3.6.5.A regarding LLS instrumentation the following two statements in the last column:

For the relief function, division specific SRV [safety relief valve] pilot SOVs

[operated safety valves] and supports will be added to the model as conservative surrogate for the LCO condition. This surrogate will have an equivalent impact as channel unavailability.

For the LLS function the SRV Fail to Reclose failure mode probability will be doubled to account for additional SRV cycles. This is conservative because additional cycles would not occur unless both trip systems were inoperable.

Accordingly, the LAR indicates that modeling for the LLS valves or its instrumentation does not currently exist in the PRA models and that SRV pilot SOVs and supporting components would need to be added before a RICT can be calculated. This infers to NRC staff that Entergy assumes in the current PRA models that there is no (or negligible) contribution from failure of the LLS valves (i.e., loss of the relief or LLS function) in accident scenarios that lead to CDF or LERF. Given this observation, it is not clear how the 30-day RICTs presented in LAR Table E1-2 for TS LCO 3.6.1.6.A and TS LCO 3.3.6.5.A were calculated.

Also, it is not clear from the LAR, how the added SRV SOVs and supporting components would be modeled to reflect the failure of both the relief and LLS function in order to calculate a RICT for TS LCO 3.6.1.6.A and TS LCO 3.3.6.5.A. More specifically, it is not clear what the impact of loss of the relief and LSS function have on safe shutdown and what the success criteria would be for losing those functions.

Given the observations above, provide the following:

i.

Explain the basis for not modeling the LLS valve relief and LLS functions in the current PRA models.

ii.

Explain how the 30-day RICTs presented in Table E1-2 for TS LCO 3.6.1.6.A and TS LCO 3.3.6.5.A were calculated.

16 iii.

Describe the modeling of SRV pilot SOVs and supports that would be added to the PRA model. Include explanation of what the impact of loss of the relief and LSS function have on safe shutdown in the PRA models and what the success criteria would be for losing those functions.

iv.

Justify that the modeling described in Part iii above is sufficient to model the as-built, as-operated plant and is consistent with the licensing basis assumptions.

g) LAR Table E1-1 states for TS LCO 3.3.5.1.F and TS LCO 3.3.5.1.G regarding ADS initiation logic and instrumentation functions that:

Division specific ADS SRV pilot SOVs and supports will be added to the ADS model as conservative surrogates for the LCO condition. This surrogate is conservative as it takes no credit for other functions.

It is not clear how the added SRV SOVs and supporting components would be modeled to reflect the failure of ADS function and whether it will be consistent with the design basis criteria.

Therefore, describe the modeling of the SRV SOVs and supporting components that would be added to reflect the failure of the ADS function in a RICT for the cited LCOs.

Include:

i.

Discussion of the impact that loss of the ADS function has on safe shutdown, ii.

Clarification of whether loss of the ADS function will be modeled consistent with the design basis criteria, and iii.

Description of what the success criteria would be for losing the ADS function.

APLA In-Scope TS/LCO Conditions RICT Estimate Notes 2, 3, and 4 of Table E1-2 of LAR enclosure 1 indicated that certain train (e.g., Standby Service Water (SSW) B)) is limiting for the RICT calculation and the calculation for the opposite train would result in longer RICT for TS LCOs 3.7.1.D, 3.8.1.B, and 3.8.1.D. The staff noted that the non-limiting train (e.g., SSW A) would have a RICT that is almost twice as long as the limiting train.

Provide justifications for the difference in RICT between the limiting train and non-limiting train for TS LCOs 3.7.1.D, 3.8.1.B, and 3.8.1.D.

APLC Exclusion of Seismic Penalty in RICT Estimates Section 2.3.1, Item 7 of NEI 06-09, 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 SE 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.

Table E1-2 of enclosure 1 of the LAR provides RICT estimates for TS actions proposed to be in

17 the scope of the RICT program. A second paragraph before the table, the licensee stated that RICTs are based on the internal events (including internal flooding), and internal fire PRA model calculations for CDF and LERF. The NRC staff notes that section 4 of enclosure 4 identifies that the seismic risk cannot be ignored to overall plant risk and calculates seismic CDF and LERF penalties that are listed in Table E5-1. It is unclear to the NRC staff whether the RICT values of Table E1-2 include the seismic penalties in the calculation.

(a) Confirm that the RICT values provided in Table E1-2 of enclosure 1 the LAR were calculated considering the seismic penalties.

(b) If seismic penalties were not provided in Table E1-2, include the seismic penalties, and update the table.

APLC Plant-Level High Confidence Lower Probability of Failure (HCLPF) Capacity Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A (ML12286A322), 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 SE for NEI 06-09 (ML071200238) states that [w]here 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 section 3 of enclosure 4 to the LAR, the licensee provided its seismic risk contribution analysis. The licensee concluded that Grand Gulf is more robust than was credited in the Generic Issue (GI)-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States, and provided the HCLPF of 0.3g and a composite uncertainty factor (c) of 0.4 as plant level fragility. The NRC staff noted that GI-199 shows HCLPF = 0.15g and c = 0.4 for Grand Gulf, which is consistent with the EPRI document, Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates, dated March 11, 2014 (ML14080A589).

The licensee provided a document to support its proposed plant-level HCLPF, GGNS-CS 00001, Revision 0, on the audit portal for NRC staff review. The NRC staff reviewed the document and identified the following questions:

(a) In GGNS-CS-23-00001, the licensee states that 0.3g PGA anchored to the ground motion response spectrum (GMRS) is a reasonable plant-level HCLPF for Grand Gulf based on experience-based screening criteria in EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, and seismic walkdowns performed as part of its IPEEE. However, the staff notes that the licensees Individual plant examination of external events (IPEEE) was based on a reduced-scope seismic margins assessment and the associated walkdowns were performed per reduced-scope walkdown guidelines. The staff considers the licensees approach to an enhanced HCLPF of 0.3g based on a reduced-scope seismic margin assessment (SMA) and associated walkdowns not acceptable without additional information and evaluation. The licensee may choose an option for a full-scope SMA to support a new HCLPF value. The licensee may upgrade its reduced-scope SMA performed as part of the Grand Gulf

18 IPEEE to a full-scope SMA consistent with applicable NRC-endorsed guidance. If this approach is adopted, the licensee must provide resulting full-scope reports, calculation notebooks, and conclusions for NRC staff review in a regulatory audit.

(b) As an alternative, the licensee may consider a scaling method based on safe shutdown earthquake to GMRS demand ratio. This method, which was approved by the staff for Waterford TSTF-505 LAR, can provide an alternative to a full-scope SMA approach discussed above in establishing an enhanced plant-level HCLPF value. In GGNS-CS-23-00001, the licensee indicates that a relatively high value of scale factor is observed for a frequency range of 1 - 10 Hz, which may result in a substantial enhancement of the HCLPF value.

(c) Re-evaluate and provide the seismic penalty for RICT calculations based on updated HCLPF and c values, if they are different from those provided in the LAR.

APLC Seismic Risk Contribution Analysis Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A (ML12286A322), 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 SE for NEI 06-09 (ML071200238) states that [w]here [probabilistic risk assessment] PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

(a) In section 3.3 of enclosure 4 to the LAR, the license states that the ratio of the internal events LERF to CDF of 0.3 was used to determine non-seismic containment failure probability. However, it is not clear to the staff how the ratio was used in computing non-seismic containment failure probability. Explain how the ratio of the internal events LERF to CDF of 0.3 was used in computing non-seismic containment failure probability which is added to the seismic contribution to determine the seismic LERF penalty.

(b) It is noted that the seismic LERF penalty of 3.58E-7 per year in enclosure 5 of the LAR is not consistent with the value of 7.16E-7 per year in section 3.3 of enclosure 4 to the LAR. Explain the different seismic LERF values of 3.58E-7/yr in enclosure 5 and 7.16E-7/yr in section 3.3 of enclosure 4 to the LAR.

(c) Section 3 of enclosure 4 to the LAR does not address the incremental risk associated with seismic-induced loss of offsite power (LOOP) that may occur following a design basis seismic event. The accident scenarios associated with seismically-induced (and therefore unrecoverable) LOOP frequency could already be addressed to some extent in the internal events PRA for unrecovered LOOP events; however, this is not explained.

Demonstrate that seismic-induced LOOP will have an inconsequential impact on RICT calculations.

APLC Extreme Wind or Tornado Screening Criteria

19 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 documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The NRC staffs SE for NEI 06-09 states that "[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk."

Table E4-1 of enclosure 4 to the LAR screens the Extreme Wind or Tornado as PS4, mean CDF is < 1E-06 per year. The description of the screening of extreme wind or tornado hazard in the table includes two discussions (1) wind loading resulting from extreme wind and tornados and (2) the Tornado Missile Risk Evaluator (TMRE) analysis for tornado missiles. NRC staff observes that the PS4 criterion applies to risk associated with tornado missiles but is not applicable to the risk associated with the wind loading from extreme wind and tornados which appear to meet the design criteria, such as C1 for Event damage is < events for which plant is designed.

(a) Justify that criterion PS4 applies to wind loading resulting from extreme wind and tornados or assign and justify a different criterion for this element of the hazard.

(b) Justify that the contribution from tornado missile CDF risk has a minimal impact on the RICT calculations showing that the impact of tornado missiles is inconsequential on the plant configuration considered in the RICT program whose RICTs are most likely to be impacted.

APLC External Flooding and Intense Precipitation As clarified in the SE on NEI 06-09, Revision 0-A, other sources of risk (i.e., seismic and other external events) must be quantitatively assessed if they contribute significantly to configuration specific risk. The SE on NEI 06-09, Revision 0-A also states that bounding analyses or other conservative quantitative evaluations are permitted where realistic PRA models are unavailable.

Table E4-1 of enclosure 4 to the LAR provides the licensees evaluation of the external flooding hazard risk. The LAR states per the flooding hazard reevaluation report that external flooding events will cause no flooding damage to Grand Gulf safety-related SSCs. The screening criteria used are (1) C1 which is Event damage is less than events for which plant is designed and (2)

PS1 which is design basis hazard cannot cause a core damage accident. NRC in review letter dated May 7, 2018 Grand Gulf Nuclear Station, Unit 1 - Staff Assessment of Flooding Focused Evaluation identifies nine doors for which installation of sandbags are credited for protection of key safety functions. The Grand Gulf emergency response procedure, 05-1-02-VI-1, Flooding lists these nine doors and states that they should remain closed for the duration of a flooding event. However, the procedure does not provide instructions concerning sandbags to protect key safety functions. Given these observations, provide the following:

(a) Provide justification for the criteria that were used to screen the external flood hazard.

Include discussion of how Entergy ensures that sandbags are installed to protect key safety functions, citing any applicable procedures and timing available for such actions.

(b) If external flood cannot be screened per the criteria cited in the LAR, and a quantitative criterion is used (e.g., PS3 or PS4), then justify that the contribution from flood risk has an inconsequential impact on the RICT calculations such as performing a sensitivity study.

20 EEEB-16 For TS 3.8.1, Condition C - Two required offsite circuits inoperable, Required Action C.2, Table A4-1, Cross-Reference of TSTF-505 and Grand Gulf Technical Specifications, of the LAR (Page 16 of 18) states (underline added)

In this Condition, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a DBA or transient.

Define sufficient onsite AC sources.

EEEB-17 The design success criteria in Table E1-1, In-scope TS/LCO Conditions to Corresponding PRA Functions, for certain TS 3.8.1 conditions, credit the Division 3 diesel generator (DG) for the AC power source by crosstie. According to the UFSAR, the Division 3 DG is dedicated to the high-pressure core spray system (HPCS). This DG supply power to the HPCS in the absence of the preferred power source (offsite power) via bus 17AC.

Provide the following:

Clarify whether the Division 3 DG has sufficient capacity and electrical distribution arrangement to supply other safety loads (ESF loads) beyond the HPCS required loads.

If it does, describe how the crosstie is modeled in PRA, specifically if manual actions are required.

EEEB-18 The design success criterion (DSC) for TS 3.8.1 condition F in Table E1-1 states With one required automatic load sequencer (Division 1 or 2) inoperable, the other Division (1 or 2) DG AND the Division 3 DG are capable of supplying associated Division 1 or 2 and 3 ESF load.

This DSC appears to address the required ESF load but does not address the required automatic load sequencer. Clarify the intent of this DSC.

STSB TS 3.8.1, Proposed RICT Placement In Attachment 2 of the LAR, the proposed change to add risk informed completion times (RICTs) for TS required action 3.8.1.A.2 (Restore required offsite circuit to OPERABLE status) is:

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR In accordance with the Risk Informed Completion Time Program AND

21 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of two divisions with no offsite power OR In accordance with the Risk Informed Completion Time Program In Attachment 2 of the LAR, the proposed change to add risk informed completion times (RICTs) for TS required action 3.8.1.B.4 (Restore required DG to OPERABLE status) is:

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from discovery of an inoperable Division 3 DG OR In accordance with the Risk Informed Completion Time Program AND 14 days OR In accordance with the Risk Informed Completion Time Program The NRC staff recognizes that the licensees proposed change is consistent with the NUREG-1434, Standard Technical Specifications - General Electric [Boiling Water Reactor]

BWR/6 Plants, TS markups in TSTF-505, Revision 2. However, it has been brought to staffs attention that some of the TSTF-505 markups contain errors, introducing potential for licensee actions to be less conservative than the original intent of the requirements. To modify completion times that include the phrase from discovery, the RICT shall start at discovery instead of the time the TS Action statement is entered, or the normal time zero. This requirement is not clear when the RICT statement is separated from the from discovery statement.

To provide clarity, revise the placement of the proposed completion times for TS Required Actions 3.8.1.A.2 and 3.8.1.B.4 similar to the placement of the proposed completion time for Required Action 3.3.5.1.F.2 (Place channel in trip), which inserts or in accordance with the Risk Informed Completion Time Program between 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and from discovery.

STSB-20 The NRC staff suggestion for licensee consideration: The proposed administrative controls for the RICT Program in TS 5.5.18 paragraph e of Attachment 2 to the LAR was based on the TS markups of TSTF-505, Revision 2. The NRC staff recognizes that the model SE for TSTF-505, Revision 2 contains improved phrasing for the administrative controls for the RICT Program in TS 5.5.7 paragraph e, namely the phrasing approved for use with this program instead of used to support this license amendment. In lieu of the original phrasing in TS 5.5.18 paragraph e, discuss whether the phrases used to support Amendment No. xxx or, as

22 discussed in the TSTF-505 model SE, approved for use with this program would provide more clarity for this paragraph.

Audit Questions for 10 CFR 50.69 LAR APLA Generic Sources of Uncertainty in EPRI TR-1026511 10 CFR 50.69 requires that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.

The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. RG 1.174, Revision 3, cites NUREG-1855, Revision 1, as acceptable guidance for addressing PRA uncertainties in risk-informed decision-making. NUREG-1855 states that this NUREG in association with EPRI TR 1016737 provide guidance on how to meet the ASME/ANS PRA standards requirements for uncertainties.

Based on LAR section 3.2.8, PRA Uncertainty Evaluations, and the NRC staffs audit of PRA Notebook PSA-GGNS-08-UNC, GGNS Impact of Model Uncertainty to 10 CFR 50.69 Categorization Process, the PRA uncertainty analysis was performed based on the process in NUREG-1855, Revision 1, EPRI 1016737, and EPRI 1026511. The licensee stated that plant specific PRA assumptions and sources of modeling uncertainty were identified from a review of the PRA notebooks and from generic sources of uncertainty in the EPRI 1016737.

However, for the PRA models supporting the LAR, it is not clear to the NRC staff how the licensee considered the generic industry sources of uncertainty for fire and level 2 events provided in EPRI 1026511.

Considering these observations, provide the following information:

a) Discuss how the generic industry sources of uncertainty for fire and level 2 events provided in EPRI 1026511 were assessed for this application consistent with NUREG-1855, Revision 1, or other NRC accepted methods. This discussion should also include the following:

(i) Summarize the assessment of these sources of uncertainty, including whether any of these sources are key PRA modeling assumptions or uncertainties for this application.

[Provide on the portal the detailed assessment that dispositions these sources of uncertainty for their impact on this application.]

(ii) For those sources of uncertainty that are key for this application, describe and provide a basis for how these key uncertainties will be addressed for SSC categorization under the 50.69 program (e.g., identify specific PRA sensitivity studies that will be performed for SSC categorization and the results of these sensitivity studies will be given to the IDP for consideration in the final risk characterization of SSCs).

23

-OR-b) Alternatively, if the generic industry sources of uncertainty for fire and level 2 events provided in EPRI 1026511 were not assessed for their impact on the PRA models supporting this application, provide a detailed justification for not assessing these uncertainties and how RG 1.174 is met for addressing PRA uncertainties in risk-informed decision-making.

APLA Dispositions of PRA Model Assumptions and Sources of Uncertainty 10 CFR 50.69 requires that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.

The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance.

NRC staff reviewed uncertainty analysis portal documents including PSA-GGNS-08-UNC, Revision 0, and identified sources of uncertainty for which further information is needed to understand their disposition for this application. Therefore, address the following:

a) Table 1, Item 5 of Entergy report PSA-GGNS-08-UNC states that the current modeling associated with HPCS and RCIC is conservative. The report states the HRA dependencies between HPCS and RCIC actions and the action to manually depressurize are conservative because these HRA combinations do not recognize the intervening success for inhibiting the Automatic Depressurization signals. The uncertainty was evaluated as not being identified a key source of uncertainty. NRC staff observes that conservatism in PRA modeling could mask SSC categorizations in the 10 CFR 50.69 program. Therefore, address the following:

i. Justify that the cited assumption has an inconsequential impact on the SSC categorization results. If the justification cannot be made qualitatively, then provide a quantitative assessment such as a sensitivity study demonstrating that the impact of this modeling uncertainty on SSC categorization results is inconsequential.

ii. If, in response to part (i) above, it cannot be determined that the cited assumption has an inconsequential impact on the SSC categorization results, then describe and provide a basis for how this uncertainty will be addressed for SSC categorization under the 10 CFR 50.69 program (e.g., identify specific PRA sensitivity studies that will be performed for SSC categorization and the results of these sensitivity studies will be given to the IDP for consideration in the final risk characterization of SSCs).

b) Table 2, Item 5 of Entergy report PSA-GGNS-08-UNC states that crediting fire wrap inappropriately could result in non-conservative risk results. Entergy report PSA-GGNS-03-FQ, Grand Gulf Nuclear Station Fire PRA Quantitation, section 5.6.3 states that fire wrap is important in Division I and Division II switchgear rooms to reduce the likelihood of having a Station Blackout due to Failures of ESF11 and ESF-21. Entergy report PSA-GGNS-03-FQ dispositions this source of uncertainty by stating the fire wrap credit

24 is in-alignment with current industry accepted approaches.

Given the stated importance of this source of uncertainty, describe the fire wrap and identify the fire protection codes to which the fire wrap conforms for the credit taken in the switchgear rooms for the fire PRA. If the conformance involves demonstration of equivalency rather than straightforward compliance, then discuss the corresponding engineering assessment.

APLA Digital Instrumentation and Control Modeling Concerning the quality of the PRA model, NEI 00-04 states that RG 1.174 and RG 1.200 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 change and the role the PRA plays in justifying the change.

Regarding digital I&C, the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist, such as the lack of industry data for digital I&C components, the difference 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 SSC categorization results.

However, it is not clear to NRC staff whether the licensee credited digital systems in the PRA models that will be used in the 50.69 program or whether this modeling can impact the SSC categorization results. Therefore, address the following:

a) Identify any digital I&C systems credited in the PRA models that will be used for categorizing SSCs under the 10 CFR 50.69 program.

b) If digital I&C systems are credited in the PRA models that will be used for categorizing SSCs under the 10 CFR 50.69 program, then justify (e.g., describe and provide the results of a sensitivity study) that the modeling uncertainty associated with crediting digital I&C systems in the PRA models has an inconsequential impact on the SSC categorization results.

-OR-Alternatively, describe and provide a basis for how this uncertainty will be addressed for SSC categorization under the 50.69 program (e.g., identify specific PRA sensitivity studies that will be performed for SSC categorization and the results of these sensitivity studies will be given to the IDP for consideration in the final risk characterization of SSCs).

APLC Alternate Seismic Approach Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant.

This includes the internal events at power PRA required by 10 CFR 50.69(c)(1)(i), as well as the risk analyses used to address external events.

25 The staff has previously requested and reviewed information to support its decision on the technical acceptability of the PRAs used in the case studies as well as details of the conduct of the case studies. This information is included in the supplements to the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, LAR for adoption of 10 CFR 50.69. The supplement to the 10 CFR 50.69 by Calvert Cliffs Nuclear Power Plant LAR dated May 10, 2019 (ML19130A180),

contained additional information related to the alternate seismic approach including incorporation by reference docketed information related to case study Plants A, C, and D; the supplement dated July 1, 2019 (ML19183A012), further clarified the information related to the alternate seismic approach (see response to RAI 4); the supplement dated July 19, 2019 (ML19200A216), provided responses to support the technical acceptability of the PRAs used for the Plant A, C, and D case studies as well as technical adequacy of certain details of the conduct of the case studies; the supplement dated August 15, 2019 (ML19217A143) clarified a response in the July 19, 2019 supplement. The supplement dated July 19, 2019, included modifications to the content of the EPRI report. In addition, the licensee removed several paragraphs related to its previous seismic submittals, categorization team evaluations, and IDP's decision process from a typical section 3.2.3.

Since the above-mentioned information was requested and reviewed by the staff for Calvert Cliffs Nuclear Power Plants LAR for adoption of 10 CFR 50.69, the staff is unable to use it for the licensees docket unless it is incorporated in the licensees LAR. The above-mentioned information is necessary for the staff to make its regulatory finding on the licensees proposed alternate seismic approach. The information is neither included in the LAR nor is it available in the EPRI report supporting the licensees proposed approach.

(a) Provide the above-mentioned information to support the staffs regulatory finding on the alternate seismic approach by either incorporating the information by referencing identified supplements or responding to the RAIs in the identified supplements for Calvert Cliffs.

(b) If differences exist between the licensees proposed alternate seismic approach and the information in the supplements stated above, identify such differences and either incorporate them in the licensees proposed approach or justify their exclusion.

(c) If the HCLPF and c values are changed as a result of Grand Gulf TSTF-505 APLC Question 02, the licensee is requested to re-evaluate and demonstrate that the seismic risk is low compared to the total plant risk to support 10 CFR 50.69 categorizations at Grand Gulf.

APLC Extreme Wind or Tornado Screening Criteria Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs.

of the LAR screens the Extreme Wind or Tornado as PS4, mean CDF is < 1E-06 per year. The description of the screening of extreme wind or tornado hazard in the table includes two discussions (1) wind loading resulting from extreme wind and tornados and (2) the TMRE analysis for tornado missiles. NRC staff observes that the PS4 criterion applies to risk associated with tornado missiles but is not applicable to the risk associated with the wind loading from extreme wind and tornados which appear to meet the design criteria, such as C1 for Event damage is < events for which plant is designed.

26 Justify that criterion PS4 applies to wind loading resulting from extreme wind and tornados or assign and justify a different criterion for this element of the hazard.

APLC External Flooding Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs.

of the LAR screens External Flooding and Intense Precipitation and states, The results of the flooding hazard reevaluation report indicate that external flooding events will cause no flooding damage to GGNS safety-related SSCs. The licensee didnt provide a list of SSCs, such as exterior doors, that are credited for this screening and must be categorized as HSS based on NEI 00-04.

Provide a list of the specific exterior doors that will be assigned HSS since they are credited for screening the external flood hazard in accordance with Figure 5-6 in NEI 00-04.