ML21134A212

Audit Meeting Agenda and Audit Questions for LAR to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, System, and Components for Nuclear Power Reactors

ML21134A212
Person / Time
Site:	Waterford
Issue date:	05/21/2021
From:	Perry Buckberg Plant Licensing Branch IV
To:	Entergy Operations
Buckberg P
References
EPID L 2020 LLA 0279
Download: ML21134A212 (16)

Text

May 21, 2021 Site Vice President Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - AUDIT MEETING AGENDA AND AUDIT QUESTIONS FOR LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEM, AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2020-LLA-0279)

Dear Sir or Madam:

By letter W3F1-2020-0047 dated December 18, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20353A433), Entergy Operations, Inc. (the licensee) submitted a license amendment request (LAR) for the Waterford Steam Electric Station, Unit 3 (Waterford 3). The proposed amendment would adopt Title 10 of the Code of Federal Regulations Section 50.69 (10 CFR 50.69), Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, into the Waterford 3 licensing basis.

On April 14, 2021 (ADAMS Accession No. ML21099A002), the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan that conveyed our intent to conduct a regulatory audit to support its review of the subject license amendment. In the audit plan, the NRC staff requested that an electronic portal be set up and provided a list of documents to be added to the portal. The NRC staff performed an initial review of the documents in the portal and developed a list of audit questions. The proposed dates for the audit are June 1, 2021, through June 3, 2021. The proposed agenda for the audit is provided as Enclosure 1 and the list of Audit questions is provided as Enclosure 2 to this letter.

If you have any questions, please contact me at (301) 415-1383 or by e-mail at Perry.Buckberg@nrc.gov.

Sincerely,

/RA Thomas J. Wengert for/

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Audit Agenda

2. Audit Questions cc: Listserv

AUDIT AGENDA REGARDING LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 Day 1 - Tuesday, June 1, 2021 9:00 AM to 12:00 PM Eastern Time (ET)

Entrance Meeting o Opening comments by U.S. Nuclear Regulatory Commission (NRC) and Entergy Operations Company, Inc. (Entergy) o Introductions and logistics Discuss APLA Audit Questions 1 and 2 1:00 PM to 4:00 PM ET Discuss APLA Audit Questions 3 and 4 Summary of the day (3:00 PM)1 Day 2 - Wednesday, June 2, 2021 9:00 AM to 12:00 PM ET Discuss APLA Audit Questions 5 and 6 1:00 PM to 4:00 PM ET Discuss APLC Audit Questions Summary of the day (4:00 PM)1 1 If discussion topics are completed early, additional discussions may include items from the next days agenda.

Enclosure 1

Day 3 - Thursday, June 3, 2021 9:00 AM to 12:00 PM ET Discuss APLA and APLC Audit Questions, as needed 1:00 PM to 4:00 PM ET Follow up on any remaining or new open action items Technical summary meeting with the Entergy audit team Formal exit meeting

AUDIT QUESTIONS TO SUPPORT THE REVIEW OF LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 By letter W3F1-2020-0047 dated December 18, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20353A433), Entergy Operations, Inc.

(Entergy, the licensee) submitted a license amendment request (LAR) for the Waterford Steam Electric Station, Unit 3 (Waterford 3). The proposed amendment would adopt 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, into the Waterford 3 licensing basis. The U.S. Nuclear Regulatory Commission (NRC) staff from the Division of Risk Assessment, Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA),

Branch B (APLB), and Branch C (APLC) have reviewed the LAR and provided the following questions to discuss during the audit.

APLA Question 01 - Open Internal Events PRA Facts and Observations Section 50.69(c)(1)(i) of 10 CFR requires that a licensees PRA must be of sufficient quality and level of detail to support the categorization process, and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

Section 50.69(b)(2)(iii) of 10 CFR requires that the results of the peer review process conducted to meet 10 CFR 50.69(c)(1)(i) criteria be submitted as part of the application.

Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009 (ADAMS Accession No. ML090410014), provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS-RA-Sa-20091 (hereafter referred to as the ASME/ANS 2009 PRA Standard) as one acceptable approach for determining the technical acceptability of the PRA.

1 American Society of Mechnical Engineers (ASME)/American Nuclear Society (ANS) PRA standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated February 2009, New York, NY.

Enclosure 2

The primary results of peer review are the facts and observations (F&Os) recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the Nuclear Energy Institute (NEI) guidance documents NEI 05-04, NEI 07-12, and NEI 12-13,2 which was accepted by the NRC.3 Section 1-A.2 of the ASME/ANS 2009 PRA Standard defines a PRA upgrade as a method as new to the PRA model, and Example 24 of the Non-Mandatory Appendix states that a new human reliability analysis (HRA) approach would constitute a PRA upgrade.

a. Attachment 3, Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items, of the LAR Enclosure presents the dispositions for two F&Os that remain open after the F&O closure review (i.e., F&Os HR-F2-01 and HR-G4-01),

which were assessed by the F&O closure review team as partially resolved based on the updates to the HRA spreadsheets. Both dispositions presented in the LAR state that the Waterford HRA was subsequently included in the use of the Electric Power Research Institute (EPRI) HRA calculator to perform the HRA. The NRC staff notes that the HRA calculator has the following HRA methods and inputs: HCR, ORE, CBDTM, PSFs, and stress levels in addition to ASEP and THERP. It is unclear to the NRC staff what HRA methods were used in both the spreadsheets and the HRA calculator.4 In light of these observations:

i. Describe the HRA methods used in the HRA spreadsheets and the HRA Calculator.

ii. Provide justification that the implementation of the HRA calculator in the Waterford 3 PRA does not constitute a PRA upgrade as defined in the ASME/ANS PRA 2009 Standard.

iii. Alternatively to Part ii, propose a mechanism to ensure a focused-scope HRA peer review is conducted on the new HRA methods and on all associated F&Os closed by the Appendix X approved process prior to implementing the categorization process.

APLA Question 02 - Open Fire PRA F&Os Section 50.69(c)(1)(i) of 10 CFR requires that a licensees PRA must be of sufficient quality and level of detail to support the categorization process, and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

Section 50.69(b)(2)(iii) of 10 CFR requires that the results of the peer review process conducted to meet 10 CFR 50.69(c)(1)(i) criteria be submitted as part of the application.

RG 1.200, Revision 2, provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS 2009 PRA Standard as one 2 Anderson, V.K., Nuclear Energy Institute, letter to Stacey Rosenberg, U.S. Nuclear Regulatory Commission, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, dated February 21, 2017 (ADAMS Accession No. ML17086A431).

3 Giitter, J., and Ross-Lee, M.J., U.S. Nuclear Regulatory Commission, letter to Krueger, G., Nuclear Energy Institute, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ADAMS Accession No. ML17079A427).

4 Table 6 of Entergy Report PSA-WF3-01-HR, Revision 3, WF3 At-Power Human Reliability Analysis, dated March 27, 2019, appears to state that the CBDTM/HCR Combination (Max) method was used.

acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the NEI guidance documents NEI 05-04, NEI 07-12, and NEI 12-13, which was accepted by the NRC in letter dated April 2017.

a. Attachment 3 of the LAR Enclosure presents the disposition for F&O FSS-C1-01, regarding the use of single versus two point heat release rate (HRR) analysis, stating this issue was evaluated by the NRC staff as not impacting the conclusions of the National Fire Protection Association (NFPA)-805 application. However, the subsequent 2017 F&O closure review team determined that the F&O remained Open since it does meet the Capability Category (CC)-II requirement but is met at CC-I using a conservative method. Given the conservative treatment, it is unclear to the NRC staff if the use of the single point treatment masks structure, system, and component (SSC) categorizations.5 In light of these observations:

i. Provide justification that the use of single HRR modeling treatment does not impact SSC categorizations.

ii. Alternatively to Part i, propose a mechanism to ensure the appropriate HRR modeling treatment is incorporated into the fire PRA model prior to implementing the categorization process.

b. Attachment 3 of the LAR Enclosure presents the dispositions for F&O FQ-C1-01, regarding the fire PRA HRA dependency analysis (DA) and remained Open following the 2017 F&O closure review team. The disposition presented in the LAR apparently states that the multiplier approach of NUREG-1921, EPRI/NRC-RES [Office of Nuclear Regulatory Research] Fire Human Reliability Analysis Guidelines, dated July 2012 (ADAMS Accession No. ML12216A104), was utilized in the fire PRA DA. It further seems to imply that further work was performed as it states, in part, that the update revision 6 Fire PRA documentation has been revised to include more detail to ensure FQ-C1 is met with more thorough and detailed documentation. It is unclear to the NRC staff what HRA dependency is currently employed in the Waterford 3 fire PRA.

The NRC staff further notes that, Section 5, Quantification, of NUREG-1921 states that the screening method as detailed in Section 5.1, Screening HRA Quantification, of NUREG-1921, which uses the multiplier method, provides a conservative human error probability (HEP) and may not be acceptable as a final value. The alternate approach, scoping described in Section 5.2, Scoping Fire HRA Quantification, of NUREG-1921 reduces those conservatisms and provides more realistic probability values.

Section 6.2, Dependency Analysis, of NUREG-1921 states that potential new dependencies can be created by fire effects on existing human failure events (HFEs).

For example, when an HFE in a third order or higher internal events DA combination is no longer viable, the dependencies are recalculated between the remaining HFEs.

NUREG-1921 concludes that the use of screening HEPs (multiplier) provide conservative DA results. The NRC staff notes that conservative treatments can mask SSC importance values from being designated high safety significant (HSS). In light of these observations:

5 Section 5.1, Internal Events Assessment, of NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005 (ADAMS Accession No. ML052910035).

i. Clarify the HRA dependency methods used in the Waterford 3 fire PRA and justify why supporting FQ-C1 is considered met by the licensee.

ii. If NUREG-1921 screening method is used, provide justification, such as a sensitivity study, that the use of the NUREG-1921 screening method does not impact any SSC categorization when compared to other realistic methods (e.g., NUREG-1921 scoping method).

iii. Alternatively to Part ii, propose a mechanism to ensure that the fire PRA HRA and DA incorporate realistic methods prior to implementing the categorization process.

APLA Question 03 - Process for Review of Key Assumptions and Sources of Uncertainty in the Internal Events PRA , Disposition of Key Assumptions/Sources of Uncertainty, of the LAR Enclosure, describes the process used for reviewing the PRA assumptions and sources of uncertainty.

The NRC staff reviewed the uncertainty documents provided on this audits electronic portal for the internal events, internal flooding, and fire PRA and found that further clarification is necessary regarding the review of assumptions and sources of uncertainty for the internal events PRA presented in Entergy Report PSA-WF3-01-QU-01. Describe the process used for reviewing the PRA key assumptions and sources of uncertainty for the application for the internal events PRA. Explain whether, and how, the plant specific assumptions and sources of uncertainty were considered during this review. Describe how the uncertainties were dispositioned for the 10 CFR 50.69 application.

APLA Question 04 - Dispositions of Key Sources of Uncertainty Sections 50.69(c)(1)(i) and 50.69(c)(1)(ii) of 10 CFR 50.69 require 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 that sensitivity studies 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 importance of components. The guidance in NEI 00-04 states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

The dispositions provided in Attachment 6 of the LAR Enclosure, for some of the key assumptions or sources of uncertainty, appear to potentially impact the SSC categorization process.

a. Item No. 3 in Table 6-2, Fire PRA Sources of Model Uncertainty, of Attachment 6 identifies fire frequencies for ignition sources as a fire PRA source of uncertainty because the Waterford 3 fire PRA utilizes the frequencies from the EPRI Supplement 1 to NUREG-6850 and credit for detection and suppression. The sensitivity study documented in the fire PRA uncertainty document, audited by the NRC staff,6 appears to show significant increases in core damage frequency (CDF) and large early release 6 Case 6 from the Entergy Report PSA-WF3-UNC-01, Revision 0 Notebook.

frequency (LERF) risks when the original NUREG-6850 values were used. However, the NRC staff notes that updated fire ignition frequencies have been published.7 Provide a detailed justification for why the ignition frequencies will not have an appreciable impact on the 10 CFR 50.69 categorization. Provide technical justification for its use and evaluate the significance of its use on the risk metrics provided in Attachment 2, Description of PRA Models Used in Categorization, of the LAR Enclosure.

b. Item No. 2 in Table 6-2 of Attachment 6 of the LAR Enclosure identifies exclusion of certain systems due to lack of cable data as a fire PRA source of uncertainty. Item No. 2 further states that the current approach used (assume equipment lacking detailed cable data is failed) will result in conservative evaluations, and [t]his conservatism would tend to result in additional SSCs being categorized as HSS in the 10 CFR 50.69 categorization process.

Describe the type of systems assumed failed in the fire PRA and provide further justification why this assumption would have a conservative impact on the 10 CFR 50.69 categorization.

c. NUREG-21958 provides an updated quantitative risk assessment method for consequential steam generator tube ruptures (C-SGTR). Section 3.1.2, CE

[Combustion Engineering] Plant Considerations, of NUREG-2195 notes that there is an increased vulnerability for CE plants due to geometry features, especially with replacement steam generators (SGs). The NRC staff notes that Waterford 3 is a CE plant, and the licensee states in the LAR that the Waterford 3 SGs have been replaced.

In Table A3-2, Open Internal Events Peer Review Findings Assessed During F&O Closure Review, of Attachment 3 of the LAR Enclosure, the disposition for internal events F&O LE4-5 states, based on a draft EPRI presentation,9 that MAAP code can accurately predict peak SG temperatures and has a minor impact on risk results. The NRC staff notes that presentation Slide No. 22 states the two code results are similar due to assumptions; however, Slide No. 40 states MAAP accuracy may be due to cancelling errors. Items Nos. 4 and 7 of Table 6-1, Internal Events/Internal Flooding PRA Assumptions & Sources of Uncertainty, of Attachment 6 of the LAR Enclosure identifies uncertainty concerns related to C-SGTR events. The disposition to Item No. 7 states that Waterford 3 uses a conservative treatment, and therefore, is more likely to result in additional HSS SSCs than to mask other results.

Discuss and provide the results on the sensitivity cases on induced SGTR on LERF and further justification that the sources of uncertainty do not impact the 10 CFR 50.69 categorization.

7 NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009, dated January 2015 (ADAMS Accession No. ML15016A069).

8 NUREG-2195, Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, dated May 2018 (ADAMS Accession No. ML18122A012).

9 Kenton, M., Erigo Technologies LLC, EPRI Perspective on Thermally-Induced Steam Generator Tube Rupture Issues, May 15, 2007 (ADAMS Accession No. ML071340053).

APLA Question 05 - Crediting of FLEX in the PRA Model NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269), provides the NRCs staff assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.200.

With regards to equipment failure probability, in the memorandum dated May 30, 2017, the NRC staff states in Conclusion 8:

The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.

With regards to HRA, NEI 16-06, Revision 0, Crediting Mitigating Strategies in Risk-Informed Decision Making, dated August 2016 (ADAMS Accession No. ML16286A297), Section 7.5, Human Reliability Assessment, recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections 7.5.4 and 7.5.5 of NEI 16-06 describe such actions to which the current HRA methods cannot be directly applied, such as: debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses; and those complex actions that require many steps over an extended period, multiple personnel and locations, evolving command and control, and extended time delays. In the memorandum dated May 30, 2017, the NRC staff states, in part, in Conclusion 11:

. . . Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, [Human Error Probabilities] HEPs associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.

In Attachment 6 of the LAR Enclosure, the licensee summarizes the credit for FLEX in the PRA.

The licensee states that the credit for FLEX equipment is limited to specific extended loss of offsite power scenarios. It further explains that only permanently installed FLEX equipment is credited, which includes a FLEX diesel generator to provide power to battery chargers and a FLEX core cooling pump to provide feedwater to the SGs.

Item No. 1 in Table 6-1 of Attachment 6 of the LAR Enclosure identifies the incorporation of FLEX strategies and equipment in the PRA model as a source of uncertainty and performed a sensitivity study that demonstrated that this model addition had an impact on station blackout risk. The results of the study10 demonstrate that the FLEX credit decreases CDF by 7 percent.

The disposition states that the inclusion of FLEX is not a source of uncertainty because it reflects the as-built, as-operated plant. The NRC staff notes the concern is in regard to the 10 Sensitivity Case No. 1 in the Entergy Report PSA-WF3-01-QU-01, Revision 2, WF3 PSA Uncertainty and Sensitivity Analysis, dated February 27, 2019.

failure probabilities for FLEX equipment and operator actions. During the audit, the NRC staff would like to discuss the following:

a. In Attachment 3 of the LAR Enclosure, the licensee states that generic failure data was judged applicable to the FLEX equipment because it is permanently installed and procedurally controlled. Discuss further the rationale for applying generic failure data to the FLEX equipment, and how the uncertainties associated with the parameter values are considered in the categorization.

b. Describe the credited operator actions related to FLEX equipment, and discuss the methodology used to assess the associated HEPs and the licensee personnel that perform these actions. The discussion should include:

i. A summary of how the licensee evaluated the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of Supporting Requirement HR-G3 of ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200.

ii. Whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that would render the equipment unavailable during an event, and whether the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200.

iii. If the licensees procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, a discussion detailing the technical bases for probability of failure to initiate mitigating strategies.

c. Based on the Waterford 3 PRA documentation audited11 by the NRC staff, it appears that the four FLEX operator actions were removed from the HRA DA due to time differences. However, the NRC staff notes that the HRA calculator Dependency Decision Tree tool designates low dependency for moderate/high stress levels independent of time or crew.

i. Provide further discussion/justification for excluding the FLEX operator actions from the HRA DA.

ii. Provide clarification as to whether the Waterford 3 HRA DA process was performed utilizing the HRA calculator tools, including the Dependency Decision Tree.

APLA Question 06 LERF Values The Waterford CDF and LERF values for each hazard (internal events, internal flooding, and internal fire) were presented in Attachment 2 of the LAR Enclosure. The NRC staff notes that when compared to the risk values provided in the safety evaluation for NFPA-80512 for Waterford 3, the reduction of LERF-to-CDF ratios for internal flooding and fire (80 percent) 11 Section 5.2 of Entergy Report PSA-WF3-01-HR, Revision 3.

12 Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Regarding Transition to a Risk-Informed Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c), dated June 27, 2016 (ADAMS Accession No ML16126A033).

appear to be significantly higher when compared to internal events (23 percent). The NRC staff understands that the internal events LERF model is usually the basis for the other PRA hazard LERF models. Therefore, the reason for the apparent significant reduction in internal flooding and fire LERF risk is unclear to the NRC staff.

a. Clarify if the updated internal events LERF model has been incorporated into both the internal flooding and fire PRA models.

b. If the internal flooding and fire PRA models utilize a different LERF model, then provide details of the differences between the three LERF hazard models.

c. Describe the reasons for the apparent significant differences in the reduction of LERF risk for internal flooding and fire when compared to internal event LERF risk.

APLC Question 01 - Alternative Seismic Approach Section 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 are adequate for the categorization of SSCs.

In the LAR, the licensee proposes to address seismic hazard risk using the alternative seismic approach for seismic Tier 1 plants described in EPRI Report 300201758313 and other qualitative considerations. The NRC staff understands that EPRI Report 3002017583 is an updated version of EPRI Report 3002012988,14 which was reviewed in conjunction with its review of the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2 LAR, for adoption of 10 CFR 50.69 dated November 28, 2018 (ADAMS Accession No. ML18333A022). The NRC staff has not endorsed EPRI Report 3002012988 as a topical report for generic use. As such, each licensee is required to perform a plant-specific review for applicability of the EPRI Tier 1 alternative seismic approach to their plant.

The NRC staff reviewed and approved the Calvert Cliffs alternative seismic approach based on the information for Tier 1 plants included in EPRI Report 3002012988, and the information provided in the supplements to the Calvert Cliffs LAR. Information in the supplements to the Calvert Cliffs LAR (ADAMS Accession Nos. ML19130A180, ML19183A012, ML19200A216, and ML19217A143) that was used to support the NRC staffs review and approval of the Calvert Cliffs alternative seismic approach is included in the NRC staffs safety evaluation for the Calvert Cliffs LAR (ADAMS Accession No. ML19330D909). The NRC staff notes that the licensees proposed alternative seismic approach is similar to that reviewed and approved in the NRC staffs Calvert Cliffs safety evaluation. However, the licensees approach for Waterford 3 is based on EPRI Report 3002017583 instead of EPRI Report 3002012988.

a. The licensee cited EPRI Report 3002017583 in the LAR; the report should be submitted on the docket for NRC staff review.

b. Explain whether the information in EPRI Report 3002012988 and in the supplements to the Calvert Cliffs LAR used to support the NRC staffs review and approval of the Calvert 13 EPRI Report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, dated February 11, 2020.

14 EPRI Report 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, dated July 2018.

Cliffs alternative seismic approach is fully represented in EPRI Report 3002017583 and the licensees LAR for Waterford 3. If there are any gaps between the two sets of information, any missing information should be identified and incorporated in the licensees LAR, as applicable.

c. Identify and justify differences, if any, between the licensees proposed alternative seismic approach and that reviewed and approved in the NRC staffs Calvert Cliffs safety evaluation, including any Waterford 3-specific considerations.

APLC Question 02 - External Hazards Screening NEI 00-04, Revision 0, provides guidance on including external events in the categorization of each SSC to be categorized. Figure 5-6 Other External Hazards, in NEI 00-04 illustrates the process that begins with the SSC selected for categorization and then proceeds through the flow chart for each external hazard. Figure 5-6 of NEI 00-04 shows that if a component participates in a screened scenario, then in order for that component to be considered as a low safety significant (LSS) item, it has to be further shown that if the component was removed, the screened scenario would not become unscreened. NEI 00-04 explicitly states, in part, that [i]f it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the low safety-significant category.

Section 3.2.4, Other External Hazards, of the LAR Enclosure states, in part, that [a]ll external hazards, except for seismic, were screened from applicability to Waterford 3 per a plant-specific evaluation in accordance with Generic Letter (GL) 88-20 [Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f), Supplement 4, dated June 28, 1991 (ADAMS Accession No. ML031150485)] and updated to use the criteria in the [ASME/ANS] PRA Standard RA-Sa-2009. Attachments 4 and 5 of the LAR provide the results of other external hazards screening and the progressive screening approach. However, the licensee does not address any considerations with respect to the application of Figure 5-6 of NEI 00-04 to external hazards screening at Waterford 3.

a. Clarify whether SSCs credited for screening of external hazards will be evaluated using the guidance illustrated in Figure 5-6 of NEI 00-04 during the implementation of the 10 CFR 50.69 categorization process at Waterford 3.

b. Identify the external hazards addressed in Attachment 4, External Hazards Screening, of the LAR Enclosure that will be evaluated according to the flowchart in Figure 5-6 of NEI 00-04.

c. If the approach illustrated in Figure 5-6 of NEI 00-04 will not be used, describe the licensees proposed approach, and provide its justification.

d. Attachment 4 to the LAR Enclosure indicates that the tornado missile hazard is screened on the basis of a recent tornado hazard analysis. It is unclear to the NRC staff if the analysis included the assessment of NRC Regulatory Issue Summary (RIS) 2015-06, Tornado Missile Protection, dated June 10, 2015 (ADAMS Accession No. ML15020A419).

i. Clarify whether the recent analysis included the RIS 2015-06 assessment.

ii. Provide justification, as applicable, that any non-conformances identified in the assessment do not impact the screening of tornado missile hazard.

iii. Alternatively to Part ii, provide an updated screening analysis for the extreme wind or tornado hazard.

APLC Question 03 - Seismic Risk Contribution Section 50.69(b)(2)(ii) of 10 CFR requires that a LAR include a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown . . are adequate for the categorization of SSCs. Section 50.69(b)(2)(iv) of 10 CFR requires that a LAR include a description of, and basis for acceptability of, the evaluations to be conducted to satisfy [10 CFR] 50.69(c)(1)(iv). The Statement of Consideration (SOC) on 10 CFR 50.69(b)(2)(iv) of the Final rule15 , states that the licensee is required to include information about the evaluations they intend to conduct to provide reasonable confidence that the potential increase in risk would be small. The SOC further clarifies that a licensee must provide sufficient information to the NRC, describing the risk sensitivity study and other evaluations and the basis for their acceptability as appropriately representing the potential increase in risk from implementation of the requirements in the rule.

In Section 3.2.3, Seismic Hazards, of the LAR Enclosure, the licensee states, in part, that low seismic CDF and LERF estimates lead to reasonable confidence that seismic risk contributions would allow reducing an HSS to LSS due to the 10 CFR 50.69 Integral Assessment if the equipment is HSS only due to seismic considerations. Section 2.2.2 of EPRI Report 3002017583 identifies the contribution of seismic to total plant risk as a basis for the use of the proposed alternative seismic approach for Tier 1 sites. However, the NRC staff notes that the LAR does not provide information to show that the plant-specific seismic risk constitutes a small fraction of the total plant risk, and thus, the proposed alternative seismic approach is applicable to Waterford 3.

In Section 3.2.3, the licensee further states that Waterford 3 completed a bounding seismic risk evaluation . . . to support development of a Risk-Informed Completion Time (TSTF-505) license amendment request and program. Based on the Technical Specifications Task Force (TSTF)

Traveler TSTF-505 LAR (ADAMS Accession No. ML21039A648) for Waterford 3, it appears that the seismic penalty was based on a high-confidence low-probability of failure (HCLPF) of 0.25g, when it appears it should be 0.1g.

a. Provide a reference where an HCLPF of 0.25g is evaluated and justified. If the value of 0.25g for the HCLPF cannot be justified, provide an updated value for the HCLPF, and make necessary changes to the estimated seismic CDF and seismic LERF values in the TSTF-505 and 10 CFR 50.69 LARs.

b. Justify that the plant-specific seismic risk is low relative to the overall plant risk, such that the categorization results will not be significantly impacted to support the applicability of the proposed alternative seismic approach to Waterford 3.

15 Final rule published in the Federal Register on November 22, 2004 (69 FR 68008).

APLC Question 04 - Change of Seismic Hazards Regulatory Position C.9, NRC Endorsement of Revision 0 of NEI 00-04; Specific Clarifications, of RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, dated May 2006 (ADAMS Accession No. ML061090627), states, in part:

As part of the NRCs review and approval of a licensees or applicants application requesting to implement §50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the licensees categorization approach. If a licensee or applicant wishes to change its categorization approach and the change is outside the bounds of the NRCs license condition (e.g., switch from a seismic margins analysis to a seismic PRA), the licensee or applicant will need to seek NRC approval, via a license amendment, of the implementation of the new approach in their categorization process.

In Section 3.2.3 of the LAR Enclosure, the licensee states:

In the unlikely event that the Waterford 3 seismic hazard changes to medium risk (i.e., Tier 2) at some future time, Waterford 3 will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

It appears this statement indicates that the licensee will switch to the Tier 2 approach, which is outside of this proposed alternative seismic approach (i.e., the Tier 1 approach), without prior review and approval by the NRC staff.

Confirm that the licensee will seek prior NRC approval if the licensees feedback process determines that a process different from the proposed alternative seismic Tier 1 approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69.

ML21134A212 *by e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DRA/APLA/BC*

NAME PBuckberg (TWengert for) PBlechman RPascarelli DATE 5/20/2021 5/19/2021 5/10/2021 OFFICE NRR/DRA/APLC* NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*

NAME SRosenberg JDixon-Herrity PBuckberg (TWengert for)

DATE 5/10/2021 5/20/2021 5/21/2021