ML22300A208

Issuance of Amendment No. 269 Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML22300A208
Person / Time
Site:	Waterford
Issue date:	11/30/2022
From:	James Drake NRC/NRR/DORL/LPL4
To:	Entergy Operations
References
EPID L-2020-LLA-0279
Download: ML22300A208 (1)

Text

November 30, 2022 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 - ISSUANCE OF AMENDMENT NO. 269 RE: ADOPTION OF 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2020-LLA-0279)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 269 to Renewed Facility Operating License (RFOL) No. NPF-38 for the Waterford Steam Electric Station, Unit 3. The amendment consists of changes to the RFOL in response to your application dated December 18, 2020, as supplemented by letters dated October 1, 2021, April 25, 2022, and August 19, 2022.

The amendment modifies the licensing basis by adding a license condition to the RFOL to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Jason J. Drake, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Amendment No. 269 to NPF-38

2. Safety Evaluation cc: Listserv

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 Renewed License No. NPF-38

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (EOI), dated December 18, 2020, as supplemented by letters dated October 1, 2021, April 25, 2022, and August 19, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. NPF-38 is hereby amended to add paragraph 2.C.22 to read as follows:

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE [Individual Plant Examination of External Events] Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergys submittal letter dated December 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 269 dated November 30, 2022.

Entergy will complete closure of the four Human Reliability Analysis (HRA) Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4 in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022, using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jennifer L. Jennifer L. Dixon-Herrity Date: 2022.11.30 Dixon-Herrity 18:57:22 -05'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-38 Date of Issuance: November 30, 2022

ATTACHMENT TO LICENSE AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 Replace the following pages of the Renewed Facility Operating License No. NPF-38 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT (a) The information in the FSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the Safety Evaluation Report Related to the License Renewal of Waterford Steam Electric Station Unit 3, are collectively the License Renewal FSAR Supplement. This Supplement is henceforth part of the FSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, EOI may make changes to the programs, activities, and commitments described in this Supplement, provided the EOI evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(b) The License Renewal FSAR Supplement, as defined in license condition 21(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

(1) EOI shall implement those new programs and enhancements to existing programs no later than 6 months before the PEO.

(2) EOI shall complete those activities by the 6 month date prior to the PEO or to the end of the last refueling outage before the PEO, whichever occurs later.

(3) EOI shall notify the NRC in writing within 30 days after having accomplished item (b)(1) above and include the status of those activities that have been or remain to be completed in item (b)(2) above.

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergy's submittal letter, dated December 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 269 dated November 30, 2022.

AMENDMENT NO. 269

Entergy will complete closure of the four Human Reliability Analysis (HRA)

Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4 in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022, using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D. The facility requires an exemption from certain requirements of Appendices E and J to 10 CFR Part 50. These exemptions are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 10 (Section 6.1.2) and Supplement No. 8 (Section 6.2.6), respectively. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

E. EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled:

"Physical Security, Safeguards Contingency and Training & Qualification Plan," and was submitted on October 4, 2004.

EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The EOI CSP was approved by License Amendment No. 234 and supplemented by a change approved by Amendment Nos. 239, 241, and 247.

F. Except as otherwise provided in the Technical Specifications or the Environmental Protection Plan, EOI shall report any violations of the requirements contained in Section 2.C of this renewed license in the following manner. Initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written follow-up within 30 days in accordance with the procedures described in 10 CFR 50.73(b), (c) and (e).

AMENDMENT NO. 269

G. Entergy Louisiana, LLC shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This renewed license is effective as the date of issuance and shall expire at midnight on December 18, 2044.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ho K. Nieh, Director Office of Nuclear Reactor Regulation

Enclosures:

1. (DELETED)

2. Attachment 2

3. Appendix A (Technical Specifications) (NUREG-1117)

4. Appendix B (Environmental Protection Plan)

5. Appendix C (Antitrust Conditions)

Date of Issuance: December 27, 2018 AMENDMENT NO. 269

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By letter dated December 18, 2020 (Reference 1), as supplemented by letters dated October 1, 2021 (Reference 2), April 25, 2022 (Reference 3), and August 19, 2022 (Reference 4), Entergy Operations, Inc. (Entergy, the licensee) submitted a license amendment request (LAR) for the Waterford Steam Electric Station, Unit 3 (Waterford 3). The licensee proposed the following license condition to the Renewed Facility Operating License (RFOL) to allow the implementation of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE [Individual Plant Examination of External Events] Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS

[American Society of Mechanical Engineers/American Nuclear Society] PRA Standard RA-Sa-2009 [Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications] for other external hazards except seismic; and the alternative seismic approach as described in Entergys submittal letter, dated December 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 269 November 30, 2022.

Enclosure 2

Entergy will complete closure of the four Human Reliability Analysis (HRA)

Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4 in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022 [ (Reference 5)], using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated and systematic risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance1.

In email correspondence dated July 26, 2021 (Reference 6), and July 22, 2022 (Reference 7),

the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff requested additional information from the licensee. The licensee responded to the requests for additional information (RAIs) in supplemental letters dated October 1, 2021, April 25, 2022, and August 19, 2022. The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 23, 2021 (86 FR 15505).

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of 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. For SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.

Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four RISC categories:

- RISC-1: Safety-related SSCs that perform safety significant functions

- RISC-2: Non-safety-related SSCs that perform safety significant functions 1 Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, May 2006, describes the SSC categorization process in its entirety as an overarching approach that includes multiple approaches and methods identified for a PRA hazard and non-PRA methods.

- RISC-3: Safety-related SSCs that perform low safety significant functions

- RISC-4: Non-safety-related SSCs that perform low safety significant functions SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.

Instead, the requirements in 10 CFR 50.69 enable licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or potentially enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on equipment that has been categorized as HSS using the requirements in 10 CFR 50.69.

2.2 Regulatory Guidance The NRC staff considered the following regulatory guidance during its review of the proposed changes:

Regulatory Guide (RG) 1.201, Revision 1 (Reference 8).

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated May 2009 (Reference 9) and RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (Reference 10).

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (Reference 11).

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, dated March 2017 (Reference 12).

NRC-Endorsed Guidance NEI issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline, dated July 2005 (Reference 13), as endorsed by RG 1.201, Revision 1, for trial use with clarifications, which describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

Sections 2 through 10 of NEI 00-04 describe the following steps/elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Section 3.2, Use of Risk Information; and section 5.1, Internal Events Assessment, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

Section 3, Assembly of Plant-Specific Inputs; section 4, System Engineering Assessment; section 5, Component safety Significance Assessment; and section 7,

Preliminary Engineering Categorization of Functions, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

Section 6, Defense-in-Depth Assessment, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

Section 8, Risk Sensitivity Study, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).

Section 2, Overview of Categorization Process, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

Section 9, IDP [Integrated Decision-making Panel] Review and Approval; and section 10, SSC Categorization, provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Additionally, section 11, Program Documentation and Change Control, of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f). Section 12, Periodic Review, of NEI 00-04 provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e). Maintaining change control and periodic review provides confidence that all aspects of the program reasonably reflect the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions on the use of PRA findings and risk insights in support of changes to a plants licensing basis, is to show that the proposed licensing basis changes meet the five key principles stated in section C of RG 1.174, Revision 3.

3.2 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles of RG 1.174, Revision 3.

3.2.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations The SSCs are classified as having either HSS functions (i.e., RISC-1 and RISC-2 categories) or LSS functions (i.e., RISC-3 and RISC-4 categories). For HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements from these SSCs). For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For RISC-3 SSCs, licensees can replace special treatment with an alternative treatment. For RISC-4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.

The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00 04, Revision 0, as endorsed in RG 1.201, Revision 1 with condition, and the acceptability of the licensees PRA for use in the application of the

10 CFR 50.69 categorization process. The NRC staffs review, as documented in this safety evaluation (SE), used the framework provided in RG 1.174, Revision 3, and NEI 00 04, Revision 0, as endorsed in RG 1.201, Revision 1.

Section 2, Overview of Categorization Process, of NEI 00-04, Revision 0, states, in part, that the categorization process includes eight primary steps. In enclosure 1 to the LAR, as supplemented by letter dated April 25, 2022, the licensee stated that it will implement the risk-informed categorization process in accordance with NEI 00 04, Revision 0, as endorsed in RG 1.201, Revision 1. In enclosure 1 to the LAR, the licensee proposed the use of the Tier 2 alternate seismic method described in Electric Power Research Institute (EPRI) report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, dated February 2020 (Reference 14), as an alternative method to assess the seismic hazard contribution(s). The method for the categorization for passive components is based on the ANO-2 methodology for passive components approved for risk-informed safety classification and treatment for repair/replacement activities in class 2 and 3 moderate- and high-energy systems (Reference 15). The NRC notes that use of these alternative methods are deviations from the NEI 00-04 guidance as endorsed. A more detailed NRC staff review of the alternative methods is provided in section 3.3.1.2 of this SE.

The licensee provided further discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 0.

The regulatory requirements in 10 CFR 50.69 and the monitoring outlined in NEI 00-04, Revision 0 and clarifications in RG 1.201, Revision 1, ensure that the SSC categorization process is sufficient to assure that the SSC functions continue to be met and that any performance deficiencies will be identified, and appropriate corrective actions will be taken. The licensees SSC categorization program includes the appropriate steps and elements prescribed in NEI 00-04, Revision 0, to assure that SSCs specified are appropriately categorized consistent with 10 CFR 50.69. The NRC staff performed a more detailed review of specific steps and elements of the licensees SSC categorization process, where necessary, to confirm consistency with the NEI 00-04 guidance, as endorsed. In light of the above, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the current regulations and therefore meets the first key principle for risk-informed decision-making prescribed in RG 1.174, Revision 3.

3.2.2 Key Principle 2: Licensing Basis Change is Consistent with the DID Philosophy In RG 1.174, Revision 3, the NRC identified the following considerations used for evaluating how the licensing basis change is maintained for the DID philosophy:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential CCFs [common-cause failures].

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

RG 1.201, Revision 1, endorses the guidance in section 6 of NEI 00-04, Revision 0, but notes that the containment isolation criteria in this section of the guidance, are separate and distinct from those set forth in 10 CFR 50.69(b)(1)(x). The criteria in 10 CFR 50.69(b)(1)(x) are to be used in determining which containment penetrations and valves may be exempted from the Type B and Type C leakage testing requirements in both Options A, Prescriptive Requirements, and B, Performance-Based Requirements, of Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to 10 CFR Part 50. The criteria provided in paragraph 50.69(b)(1)(x) of 10 CFR are not to determine the proper RISC category for containment isolation valves or penetrations.

In enclosure 1 to the LAR, as supplemented by letter dated April 25, 2022, the licensee clarified that the DID assessment will be performed in accordance with NEI 00-04, Revision 0. Based on the above, the NRC staff concludes that the proposed change is consistent with the DID philosophy described in Key Principle 2 of RG 1.174, Revision 3, and is, therefore, acceptable.

The NRC staff finds that the licensees process meets the requirements in 10 CFR 50.69(c)(1)(iii) that requires DID to be maintained.

3.2.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The regulations in 10 CFR 50.69(c)(1)(iv) require the evaluations to provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF), resulting from changes in treatment, are small. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of the SSC to assure that sufficient safety margins are maintained. With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the licensing basis (e.g., Final Safety Analysis Report (FSAR), supporting analyses) are met, or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data. RG 1.174, Revision 3 provides guidelines for making that assessment including evaluations to ensure the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The SSCs design basis function as described in the plants licensing basis, including the updated FSAR and Technical Specifications Bases do not change and should continue to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the NRC staff concludes that the licensees categorization program ensures sufficient safety margins are maintained in accordance with the third key principle of RG 1.174, Revision 3, and would therefore meet the requirements set forth in 10 CFR 50.69(c)(1)(iv).

3.3 Risk-Informed Assessment 3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04, Revision 0, endorsed by RG 1.201, Revision 1, addresses the fourth and fifth key principles of the NRC staffs guidance for risk-informed decision-making, pertaining to the assessment for change in risk and monitoring the impact of the licensing basis change.

A summary of how the licensees SSC categorization process is consistent with the guidance and methodology prescribed in NEI 00-04, Revision 0, and RG 1.201, Revision 1 is provided in the sections below:

In sections 3.2.1, Internal Events and Internal Flooding, and 3.2.2, Internal Fire Hazards, of the enclosure to the LAR, the licensee described that the Waterford 3 categorization process uses PRA modeled hazards to assess risks for the internal events (including internal floods) and internal fires. For the other risk contributors, the licensees process uses the following non-PRA methods to characterize the risk:

Seismic Hazard: Tier 2 alternate method provided in EPRI report 3002017583.

External Hazards and Other Hazards: Screening analysis performed for IPEEE (Reference 16) updated using criteria from Part 6 of the ASME/ANS RA Sa-2009, Addendum A to RA S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, (the PRA Standard), as endorsed by the NRC in RG 1.200 (Reference 17).

Shutdown Events: Safe Shutdown Risk Management program consistent with Nuclear Management and Resources Council, Inc. (NUMARC) 91-06, Guidelines for Industry Actions to Assess Shutdown Management, dated December 1991 (Reference 18).

Passive Components: ANO 2 passive categorization methodology (Reference 15).

The approaches and methods proposed by the licensee to address internal events, fire, external events, other hazards, and shutdown events are consistent with the approaches and methods included in the guidance in NEI 00-04, Revision 0. The method for the categorization for passive components is based on the ANO-2 methodology for passive components approved for risk-informed safety classification and treatment for repair/replacement activities in class 2 and 3 moderate- and high-energy systems. The use of the ANO-2 methodology in the SSC categorization process is addressed in section 3.3.1.3 of this SE. To address seismic hazard in the SSC categorization process, the licensee proposed to use an alternative method not endorsed by the NRC in NEI 00-04. A detailed NRC staff review of the licensees proposed alternative seismic approach is provided in section 3.3.1.2.1 of this SE.

3.3.1.1 Scope of the PRA The NRC staff reviewed two aspects of the PRA with regard to the impact of the proposed changes on plant operational risk: (1) scope and acceptability of the PRA models and their application to the proposed changes, and (2) a review of the PRA results and insights described in the licensees application.

Evaluation of PRA models The licensee proposed PRA models to evaluate risk associated with internal events, including internal flooding and fire PRA. In section 3.2, Technical Adequacy Evaluation, of the enclosure to the LAR, the licensee confirmed that the PRA models had been peer reviewed and assessed against RG 1.200. The internal events (including internal flooding) PRA (IEPRA) was peer reviewed using the ASME/ANS RA-Sb-2005 PRA Standard as endorsed by RG 1.200 Revision 1. Further the IEPRA and fire PRA (FPRA) were peer reviewed using the ASME/ANS RA-Sa-2009 PRA Standard as endorsed by RG 1.200, Revision 2. For the IEPRA, this included focused-scope peer reviews on internal flooding, LERF, and HRA. The licensee stated it conducted an independent assessment process for closure of the finding-level F&Os resulting from these peer reviews. The NRC staff confirmed that the licensee performed closure of the F&Os using the NRC-accepted process documented in the NEI letter to the NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, dated February 21, 2017 (Reference 19), as endorsed in RG 1.200, Revision 3. In attachment 3 of the supplement dated April 25, 2022, the licensee stated that a focused-scope peer review was conducted in December 2021 on the human reliability supporting requirements that resulted in four open F&Os. In table A3-2 of the April 25, 2022, supplement, the licensee provided dispositions of all remaining open F&Os on this application. In the supplement dated August 19, 2022, the licensee proposed a license condition to complete closure of the open human reliability F&Os using the NRC accepted process. The NRC staff reviewed the remaining open F&Os and concluded they have no impact on the application. The license condition is discussed in section 4.0 of this SE.

The NRC staff determined that the Waterford 3 FPRA model was adequately updated for this application. However, with regards to the FPRA fire ignition frequencies not implementing the updated industry data in 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 (Reference 20), the licensee stated in response to Probabilistic Risk Assessment Licensing Branch A (APLA) RAI 03.a, by supplemental letter dated October 1, 2021, that several system categorizations were impacted by this issue. The licensee provided a commitment in enclosure 2 of the April 25, 2022, supplement, to update the frequencies by December 2023. The licensee further provided a commitment to provide the results of the sensitivity study to the IDP if any of the systems for which the sensitivity study shows changes in the safety significance in the FPRA model are selected for categorization prior to the update of the ignition frequencies to the industry consensus approach. Based on its review, the NRC staff finds that this approach is consistent with the approved NEI 00-04 guidance.

In response to APLA RAI 03.b, by supplemental letter dated October 1, 2021, regarding unknown cable locations for fire impacts, the licensee provided the results of a sensitivity study that demonstrated this source of uncertainty did not impact categorization results. The NRC staff finds that the assessment performed to identify the key assumptions/sources of uncertainty is consistent with the guidance provided in NUREG-1855, Revision 1.

Credit for Diverse and Flexible Mitigation Capability (FLEX) Equipment The 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 (Reference 21), provides the NRC staffs assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in

support of risk-informed decision making in accordance with the guidance of RG 1.200, Revision 2.

In its application, the licensee initially stated that while the Waterford 3 PRA credits FLEX equipment or FLEX strategies in the PRA, the PRA does not constitute as a source of uncertainty since it represents the as-built, as-operated plant. However, the NRC staff noted that the focus of the NRC staff memorandum dated May 30, 2017, is on the uncertainty of the methods used to model FLEX equipment and operator actions. Therefore, the NRC staff requested further details in APLA RAI 04.a that the uncertainty related to FLEX does not impact categorization. The licensee stated in its response by supplemental letter dated October 1, 2021, that upon further analysis it determined that FLEX modeling is a key source of uncertainty, as defined in NUREG-1855. The licensee stated that the results of the no FLEX credit sensitivity study will be provided to the IDP for their final categorization determination. The licensee documented its commitment to this effort in enclosure 2 of its supplement dated April 25, 2022. The NRC staff determined this approach is consistent with the approved NEI 00-04 guidance.

In response to NRC questions of other FLEX modeling attributes that the licensee determined to be currently excluded from the PRA model, regarding FLEX pre-initiators (APLA RAI 04.b.ii(a)),

extended loss of AC power (ELAP) operator declaration (APLA RAI 04.b.iii(a)) and including the FLEX human failure events into the HRA dependency analysis (APLA RAI 04.c.i), the licensee stated that these exclusions would be addressed in the next scheduled PRA model update, which is likely to occur after implementing the categorization program. Given that the licensee subsequently classified FLEX as a key source of uncertainty for this application, the NRC staff determined this approach is consistent with the approved NEI 00-04 guidance.

Therefore, the NRC staff concludes that the licensees treatment of FLEX strategies is acceptable for this application because the FLEX modeling uncertainty is identified as a key source of uncertainty and will be presented to the IDP.

3.3.1.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization As part of its proposed integrated decision-making process to categorize SSCs according to safety significance, the licensee has proposed to use a non-PRA method to consider seismic hazards. Sections 50.69(c)(1)(ii) and 50.69(b)(2)(ii) of 10 CFR permits the use of non-PRA methods in a risk-informed categorization process.

3.3.1.2.1 Seismic Approach In the LAR, the licensee proposed an alternative seismic Tier 1 approach, which was based on the guidance in EPRI report 3002017583. The seismic Tier 1 approach applies to plants with low seismic hazards and low relative seismic risk contribution to the total plant risk. However, the NRC staffs review of the LAR and the licensees response to Probabilistic Risk Assessment Licensing Branch C (APLC) RAI 03 dated October 1, 2021, indicated that, although seismic CDF estimate is low relative to the total plant CDF (approximately 15 percent), the relative contribution of seismic LERF is approximately 89 percent of the total plant LERF, which is not sufficiently low to justify the applicability of the Tier 1 approach for Waterford 3. These estimates for seismic CDF and seismic LERF are based on the latest publicly available information about the plant-level high-confidence of low-probability of failure (HCLPF) value of 0.15g. Even if the NRC staffs estimation of seismic LERF were to use a higher containment HCLPF value of 0.3g (based on appendix B of NUREG/CR-4334, An Approach to the

Quantification of Seismic Margins in Nuclear Power Plants, dated August 1985 (Reference 22)), the relative contribution of seismic LERF would be 70 percent of the total plant LERF, which continues to not be sufficiently low to justify the applicability of the Tier 1 approach for Waterford 3. Subsequently, to address the NRC staffs concern, the licensee proposed an alternative seismic Tier 2 approach in its supplement dated April 25, 2022. The licensees alternative seismic Tier 2 approach has two important bases: (1) the impact of the seismic risk in categorization due to the high relative contribution of seismic risk to the overall plant risk, and (2) the conclusions from the case studies in EPRI report 3002017583.

In its supplement dated April 25, 2022, the licensee stated that its basis for the seismic Tier 2 approach is that consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. The licensee explained that the basis for using the proposed alternative seismic approach is that the special seismic risk evaluation process for the proposed approach can identify the appropriate seismic insights to be considered with other categorization insights by the IDP for the final HSS determinations. The licensee stated that Waterford 3s proposed alternative seismic approach follows the same approach as approved for LaSalle County Station, Units 1 and 2 (LaSalle) in License Amendments Nos 249 and 235, respectively, dated May 27, 2021 (Reference 23), thereby making LaSalles approach the precedent for its proposed alternative seismic approach. The licensee further stated that its proposed Tier 2 approach is specified in EPRI report 3002017583 with associated revision markups provided in the LaSalle LAR supplements dated October 16, 2020 (Reference 24), and January 22, 2021 (Reference 25), which Waterford 3 incorporated by reference into its LAR.

To capture the potential impact of seismic risk in the categorization process, the licensees alternative seismic approach includes both quantitative and qualitative assessments of plant SSC-specific seismic insights and their presentation to the IDP as part of its decision-making.

The proposed Tier 2 approach includes focused walkdowns and quantification of PRA importance measures, based on a surrogate sensitivity study for selected SSCs using the licensees IEPRA. The proposed approach also includes consideration of seismic risk through insights from other plant-specific seismic information.

Summary of Case Studies in EPRI Report 3002017583 EPRI report 3002017583 includes the results from case studies performed to determine the extent and type of unique HSS SSCs from seismic PRAs (SPRAs). The case studies were performed for four plants, designated as plants A through D in the report. Description and evaluation of these case studies were documented in the NRC staffs SE for LaSalles 10 CFR 50.69 approval, which the licensee identified as the precedent for the licensees proposed alternative seismic approach.

Evaluation of the Information Provided for the Proposed Alternative Seismic Approach In section 3.2.3, Seismic Hazards, of enclosure 1 to the April 25, 2022, supplement, the licensee provided a description of its proposed Tier 2 alternative seismic approach for considering seismic risk in the categorization process and how the proposed alternative seismic approach would be used in the categorization process. The licensee cited LaSalles 10 CFR 50.69 approval as the precedent for its proposed alternative seismic approach and indicated that its proposed approach followed the approach taken in the LaSalle 50.69 amendment. In addition, the licensee based the acceptability of its proposed alternative seismic approach on the conclusions gained from the case studies in EPRI report 3002017583 and, therefore, indirectly, on the acceptability of the PRAs used for the case studies.

The information presented in the LAR, as supplemented, as well as that in the EPRI report, taken together, provides sufficient details for the licensees proposed alternative seismic approach for Waterford 3, including how the licensees proposed alternative seismic approach would be used in the categorization process, and the measures for assuring that the quality and level of detail for the licensees proposed alternative seismic approach are adequate for the categorization of SSCs. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(ii) are met for Waterfords proposed alternative seismic approach.

The information presented in the April 25, 2022, supplement, along with the documents that the licensee incorporated by reference in its application, provides sufficient description and basis for acceptability of the evaluations to be conducted to satisfy 10 CFR 50.69(b)(1)(iv) for the proposed Tier 2 alternative seismic approach. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed alternative seismic approach at Waterford 3 (i.e., the Tier 2 alternative seismic approach).

Evaluation of Technical Acceptability of the PRAs Used for Case Studies Supporting the Proposed Alternative Seismic Approach In its April 25, 2022, supplement, the licensee provided information concerning the case studies, mapping approach, and conclusions on the determination of unique HSS SSCs from the case studies which were used by the licensee to support its proposed alternative seismic approach.

The key categorization conclusion from the plants A, C, and D case studies is that the only SSCs identified as HSS in the SPRA that were not also HSS from IEPRA and/or FPRA were from unique seismically induced failure modes. The remainder of HSS SSCs from SPRA are captured by the corresponding IEPRA and/or FPRA or other aspects of the NEI 00-04, Revision 0, categorization process.

The licensee stated that it was using the case study (termed test case by the licensee) information in EPRI report 3002017583. The licensee also incorporated by reference in its application information related to technical acceptability of the PRAs used, as well as the technical adequacy of certain technical details of the conduct of the case studies, for case study plants A, C, and D. The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the case studies for Plants A, C, and D in EPRI report 3002017583 for this application. The NRC staff also evaluated the peer review process and resolution of peer-review findings, and key assumptions and sources of uncertainty for plants A, C, and D, which was incorporated by reference by the licensee.

Based on the above, the NRC staff finds that the technical acceptability of PRAs used for the case studies for plants A, C, and D in EPRI report 3002017583, the mapping approach used in those case studies, and the conclusions on the determination of unique HSS SSCs from the case studies in the LaSalle alternative seismic approach are applicable to this licensees proposed plant-specific alternative seismic approach. Therefore, the NRC staff concludes that the PRAs for plants A, C, and D were technically acceptable and applicable for use in the corresponding case studies supporting the licensees proposed alternative seismic approach and that the mapping of SSCs between the SPRA, the full-power IEPRA, and, as applicable, the FPRA for the case studies for plants A, C, and D are applicable to the licensees proposed alternative seismic approach. The licensees plant-specific evaluation is technically justifiable to support conclusions on the determination of unique HSS SSCs from SPRAs in case studies for plants A, C, and D in the EPRI report 3002017583, and applicable to Waterford 3 and the licensees proposed Tier 2 alternative seismic approach.

Evaluation of the Implementation of the Proposed Alternative Seismic Approach The categorization conclusions from the case studies in EPRI report 3002017583 indicated that seismic-specific failure modes resulted in HSS categorization uniquely from SPRAs. Therefore, such seismic-specific failure modes, such as correlated failures, interaction failures, relay-chatter, and passive component structural failure mode, can influence the categorization process. The licensee discussed the implementation of its proposed Tier 2 alternative seismic approach in its April 25, 2022, supplement. The proposed alternative seismic approach includes a combination of qualitative and quantitative considerations of the mitigation capabilities as well as seismic failure modes of SSCs in the categorization process. These considerations are based on plant-specific walkdowns for the SSCs undergoing categorization, quantification of the impact of seismic failure of SSCs subject to correlated or interaction failures, and insights obtained from prior seismic evaluations performed for Waterford 3.

Qualitative Evaluation for the Alternative Seismic Approach The licensee stated that as part of the categorization teams preparation of the system categorization document (SCD) that is presented to the IDP, a section will be included which summarizes the identified plant-specific seismic insights pertinent to the SSC being categorized.

The licensee further explained that at several steps of the categorization process, the categorization team will consider the available seismic insights relative to the system being categorized and document their conclusions in the SCD. In addition, the IDP would be provided with the basis for the proposed alternative seismic approach including the seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach.

Section 3.1.1, Overall Categorization Process, of enclosure 1 to the April 25, 2022, supplement included an explicit mention of the categorization evaluation for seismic hazard which would be performed, at either the function level, component level, or both, using the proposed alternative seismic approach.

In section 3.2.3 of enclosure 1 to the April 25, 2022, supplement, the licensee explained that the categorization team would review available Waterford 3 plant-specific seismic information and other resources to identify plant-specific seismic insights relevant to the SSCs being categorized such as:

• Impact of relay-chatter

• Implications related to potential seismic interactions such as with block walls

• Seismic failures of passive SSCs such as tanks and heat exchangers

• Any known structural or anchorage issues with a particular SSC

• Components implicitly part of PRA-modeled functions (including relays)

The licensee stated that, for each system categorized, the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs. The licensee further explained that these insights would provide the IDP a means to consider potential impacts of seismic events in the categorization process. The licensee stated that the IDP could challenge, from a seismic perspective, any candidate LSS recommendation for any SSC if they believed

there was basis for doing so, and that any decision by the IDP to downgrade preliminary HSS components to LSS would also consider the applicable seismic insights.

The licensee explained that sources of the insights related to seismic events would be prior plant specific seismic evaluations such as the seismic hazard screening, spent fuel pool assessment, expedited seismic evaluation process, as well as the seismic high frequency evaluation performed for Fukushima Near-Term Task Force (NTTF) Recommendation 2.1, seismic walkdowns performed for NTTF Recommendation 2.3, and seismic mitigation strategy assessment performed for NTTF Recommendation 4.2.

In its April 25, 2022, supplement, the licensee stated that for SSCs that were uniquely HSS from the FPRA but not HSS from IEPRA, the categorization team would review design-basis functions of the SSC(s) during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events. The results of the review would be presented to the IDP as additional qualitative inputs and would be described in the SCD. The licensee further clarified that the discussion with the IDP will focus on SSCs that are uniquely HSS from FPRA because such SSCs may not be categorized as HSS following the integrated importance measure determination.

The NRC staff concludes, based on its review of the qualitative evaluations for seismic risk in the licensees proposed alternative seismic approach, that: (1) the evaluations will include potentially important seismically induced failure modes, as well as mitigation capabilities of SSCs during seismically induced design basis and severe accident events consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI report 3002017583, (2) the licensee will provide system-specific qualitative seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, (3) the insights will use plant-specific prior seismic evaluations, which, in conjunction with the performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration, and (4) the qualitative evaluation will complement focused walkdowns and quantitative evaluations identified for the SSCs. Further, the recommendation for categorizing civil structures in the proposed Tier 2 alternative seismic approach provides appropriate consideration of such failures from a seismic event.

Focused Walkdowns for the Alternative Seismic Approach In section 3.2.3 of enclosure 1 to the April 25, 2022, supplement, the licensee stated that the proposed alternative seismic approach includes focused walkdowns of SSCs undergoing categorization. The purpose of the walkdowns is to identify, for the SSCs that are being categorized, the conditions for occurrence of correlated failures, failure of more than one SSC due to interactions with other SSCs, and single component failures.

The NRC staff evaluated the focused walkdowns for the proposed alternative seismic approach, as described in the LAR and its supplements, and in EPRI report 3002017583 including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements dated October 16, 2020, and January 22, 2021, that are incorporated by reference by Waterford 3 licensee in its application.

The licensee cited LaSalles 10 CFR 50.69 approval as the precedent for its proposed alternative seismic approach and stated that its proposed approach followed that of the precedent. The NRC staffs review of the April 25, 2022, supplement and the documents incorporated by reference in the supplement determined that the NRC staff evaluation of focused walkdowns documented in the LaSalle 10 CFR 50.69 SE is applicable to Waterford 3.

The NRC staffs review of the focused walkdowns in the proposed alternative seismic approach described in the April 25, 2022, supplement and in EPRI report 3002017583, including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements that are incorporated by this licensee in its application, finds that:

1. The licensees focused walkdown in the proposed alternative seismic approach:

(i) includes consideration of seismically induced correlated and interaction failures that fail more than one SSC as well as single component failures, (ii) includes evaluations of the direct and indirect impacts of seismically induced correlated and interaction failure of an SSC, (iii) shows that these failure modes reflect the insights from the case studies in the EPRI report, and (iv) shows the modifications to the proposed alternative seismic approach through changes to the EPRI report 3002017583 appropriately reflect the evaluation of such direct and indirect impacts.

2. The qualification of personnel performing the walkdowns and the documentation as well as retention of the walkdown results is acceptable for the proposed alternative seismic approach. The qualification of personnel performing the walkdowns for the proposed alternative seismic approach is consistent with the state-of-practice for development and peer review of contemporary SPRAs, and the documentation and retention of walkdown information for the proposed alternative seismic approach is consistent with state-of-practice SPRAs and that the guidance in NEI 00-04 will result in appropriate information being presented to the IDP for categorization decisions.

3. The licensees approach for selecting the screening criterion is consistent with that for state-of-practice SPRAs and that SSCs screened out based on the criterion are not expected to result in HSS components within the 10 CFR 50.69 categorization process.

4. The fragility approaches proposed for development of fragility values in Step 5b of the proposed alternative seismic approach are acceptable for the proposed alternative seismic approach because (i) they represent state-of-practice approaches consistent with those used in contemporary SPRAs reviewed by the NRC staff, and (ii) no unreviewed methods would be used for fragility calculations.

5. The personnel performing fragility evaluations for the proposed alternative seismic approach will have experience or background consistent with that used for state-of-practice SPRAs as well as the guidance in NEI 00-04 on personnel qualifications, and the use of such personnel is, therefore, acceptable for the proposed alternative seismic approach. In addition, the NRC staff review determined that the documentation of the fragility evaluations will be consistent with documentation used for other categorization processes and is therefore acceptable for the proposed alternative seismic approach.

6. The proposed alternative seismic approach will result in consideration of relays as implicitly modeled components and of insights related to the impact of seismically induced relay-chatter for the function achieved by the SSC during the categorization.

7. The focused walkdowns of SSCs undergoing categorization will identify seismic interaction and correlated failures including those resulting from potential failures of passive components as well as structural and anchorage issues. Further, the NRC staff concludes that insights from available plant-specific seismic reviews will also provide categorization-related insights from a seismic failure modes perspective.

Quantitative Evaluation for the Alternative Seismic Approach In section 3.2.3 of enclosure 1 to the April 25, 2022, supplement, the licensee explained that SSCs identified as being vulnerable to correlated or interaction failure modes based on the walkdown would be subjected to a quantitative evaluation using the licensees IEPRA to determine the impact of seismic events on the categorization. The quantitative evaluation would be performed through a sensitivity study, termed the surrogate sensitivity, using the licensees IEPRA. The NRC staff noted that further details on the surrogate sensitivity are provided in section 2.3.1, Description of the Approach, of EPRI report 3002017583, including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements that are incorporated by reference by the licensee in its application. The surrogate sensitivity would be performed by introducing PRA basic events, termed surrogate events, in the licensees IEPRA at appropriate locations to reflect seismically induced correlated failure or interaction failure of single or multiple SSCs.

Subsequently, the modified IEPRA with the surrogate events would be quantified for the loss-of-offsite power (LOOP) and small break loss-of-coolant accident (small LOCA) initiators and importance measures would be derived. The importance measures for the surrogate events derived from this sensitivity study would be used to identify the SSCs that should be HSS due to seismically correlated failures or seismic interaction related failures. The licensee further stated that the quantitative evaluation to determine the importance of SSCs on a system basis in the proposed alternative seismic approach was detailed in section 2.3.1 of EPRI report 3002017583.

The NRC staff reviewed the quantitative evaluation for the alternative seismic approach described in the licensees April 25, 2022, supplement and in EPRI report 3002017583, including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements, and determined that the previous NRC staff quantitative evaluation documented in the LaSalle 10 CFR 50.69 SE is applicable to Waterford 3.

The NRC staff determined that seismically induced LOOP and small LOCA occurrence frequencies are representative for Waterford 3 based on the three SPRAs in the case studies in EPRI report 3002017583 and the fact that the seismic hazard at the licensees site is lower than the hazard for those SPRAs. Therefore, the NRC staff concludes that the proposed occurrence frequency for the seismically induced LOOP event of 1.0 per year, the proposed occurrence frequency for the seismically induced small LOCA event of 1.0E-2 per year, and the proposed surrogate event failure probability of 1.0E-4 are acceptable for use in the licensees alternative seismic approach. Further, the NRC staff determined that the occurrence frequency and failure probability switch in the surrogate sensitivity is acceptable for the licensees alternative seismic approach because: (1) it is necessary for developing the importance measures for comparison against the corresponding thresholds in NEI 00-04, and (2) it does not alter the basis for the proposed values. Based on its review, the NRC staff finds reasonable confidence that the categorization outcome from the licensees proposed alternative seismic approach will be comparable to those from SPRAs.

Conclusions on the Implementation of the Alternative Seismic Approach Based on its review of the proposed alternative seismic approach for Waterford 3, in conjunction with requirements in 10 CFR 50.69 and the corresponding Statement of Consideration (SoC)

(69 FR 68007; November 22, 2004), the NRC staff finds that the proposed alternative seismic approach provides reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(ii) and (iv) and meets the intent of the SoC because:

1. It includes qualitative consideration of seismic events at several steps of the categorization process including documentation of the information for presentation to the IDP as part of the integrated, systematic process for categorization.

2. It includes focused walkdown(s) which evaluate(s) the direct and indirect impacts of seismically induced correlated failures, interaction failures, and single component failures in a system under categorization.

3. It includes a quantitative evaluation, with justified failure probability and initiating event frequencies, that provides reasonable confidence that the categorization results from the licensees proposed alternative seismic approach will be similar to those from SPRAs.

4. Personnel performing necessary walkdowns and analyses will have qualifications consistent with the state-of-practice SPRAs and the guidance in NEI 00-04. The documentation of these walkdowns and analyses will be consistent with state-of-practice SPRAs and the guidance in NEI 00-04.

5. The quantitative and qualitative insights presented to the IDP include potentially important seismically induced failure modes as well as mitigation capabilities of SSCs during seismically induced design basis and severe accident events, consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI report 3002017583 with the markups provided in the LaSalle 10 CFR 50.69 LAR supplements, which were incorporated by reference by the licensee in this application.

The quantification will use the licensees IEPRA, and the insights will use prior plant specific seismic evaluations. Therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, the proposed alternative seismic approach will reasonably reflect the current plant configuration.

6. It presents system-specific insights and categorization results from a seismic risk perspective to the IDP for consideration as part of the IDP review process, thereby providing the IDP with a means to consider potential impacts of seismic events in the categorization process.

7. It presents the IDP with the basis for the proposed alternative seismic approach including the moderate seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach.

Evaluation for Performance Monitoring for the Alternative Seismic Approach In section 3.5, Feedback and Adjustment Process, of the enclosure to the April 25, 2022, LAR supplement, the licensee stated that its configuration control process ensured that changes to the plant, including a physical change and changes to documents, are evaluated to determine the impact on design bases, licensing documents, programs, procedures, and training.

The NRC staff evaluated the licensees discussion of its performance monitoring program for the proposed alternative seismic approach to ensure: (1) the continued validity of the plant-specific information that were developed for each SSC that is categorized, (2) that any changes to the plant, including the seismic hazard, are captured and appropriately addressed as part of the 10 CFR 50.69 program, and (3) that the requirements in 10 CFR 50.69(e) were met for the proposed alternative seismic approach.

In section 3.5, Feedback and Adjustment Process, of enclosure 1 to the April 25, 2022, supplement, the licensee stated that its performance monitoring process requires periodic review to assess changes that could impact the categorization results and to provide the IDP with an opportunity to recommend categorization and treatment adjustments due to such changes. The licensee explained that its configuration control program had been updated to have a checklist related to the impact of seismic events on categorization. The licensee identified some of the items in the checklist in section 3.5 of enclosure 1 to the April 25, 2022, supplement.

The licensee stated that its performance monitoring program required that SCDs could not be approved by the IDP until the panels comments on issues, including system-specific seismic insights, had been resolved to the satisfaction of the IDP.

The licensee explained that its scheduled periodic reviews would occur no longer than once every two refueling outages and would evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it was determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process would be updated. The licensee explained that if a PRA model or other risk information is updated, a review of the SSC categorization would be performed in addition to the periodic review.

The NRC staff recognizes that the seismic hazard at any site could potentially increase such that the categorization process may be impacted from a seismic risk perspective, either solely due to the seismic risk or via the integrated importance measure determination. In section 3.2.3 of enclosure 1 to the April 25, 2022, supplement, the licensee stated that if the Waterford 3 seismic hazard changed at some future time and if its feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration under 10 CFR 50.69, it will seek prior NRC approval for use of such an approach.

The NRC staff notes that seeking prior NRC approval for use of a process different from the proposed alternative seismic approach is consistent with the license condition proposed by the licensee in section 2.3, Description of the Proposed Change, of enclosure 1 to the April 25, 2022, supplement. The licensee further stated that after receiving NRC approval, it would follow its categorization review and adjustment process procedures and would update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

Based on its review, the NRC staff finds that the licensees configuration control program includes consideration of seismic issues as well as failure modes such as interaction between components and review of seismic loading and seismic dynamic qualification. Further, the licensees performance monitoring program assesses changes that impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments due to such changes. Therefore, the NRC staff finds that the licensees performance monitoring and configuration control process addresses plant-specific seismic

evaluation, thereby ensuring that the corresponding impacts on SSC categorization continues to remain valid and if necessary, are presented to the IDP for consideration of categorization changes.

During its review, the NRC staff noted that the licensees performance monitoring program for 10 CFR 50.69 has the capability to identify significant changes to the plant risk profile as well as instances in which a RISC-3 or RISC-4 SSC may fail to perform a safety significant function, resulting in an immediate evaluation and review for such instances. Based on its review, the NRC staff finds that the requirements in 10 CFR 50.69(e) are met for the proposed alternative seismic approach.

Conclusion for Proposed Alternative Seismic Approach Based on its review, the NRC staff concludes that the licensees proposed alternative seismic approach for Waterford 3, as described in the licensees April 25, 2022, supplement is acceptable for considering seismic risk in the licensees categorization process under 10 CFR 50.69.

3.3.1.2.2 Other Non-Seismic External Hazards This hazard category includes all non-seismic external hazards such as high winds, external floods, transportation, nearby facility accidents, and other hazards. In Entergy Report, PSA-WF3-07-01, Waterford 3 Re-Examination of External Events Evaluation in the IPEEE, dated November 2017, the licensee states, in part, [t]he high winds, floods, transportation and other external events (HFO) areas were eliminated based on either compliance with 1975 NRC Standard Review Plan (SRP) criteria or on the basis of a bounding probabilistic assessment resulting in a CDF estimate less than 1E-6 per reactor year, i.e., below the NUREG-1407,

[Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities] screening criterion.

In section 3.2.4, Other External Hazards, of the enclosure to the LAR, the licensee states, in part, that all other 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 for Severe Accident Vulnerabilities - 10 CFR 50.54(f)] and updated to use the criteria in [ASME/ANS] PRA Standard RA-Sa-2009. In the October 1, 2021, supplement, the licensee stated in response to RAI APLC 02 that Waterford 3 will subject all of the external hazards (excluding internal fires and seismic hazards) to the process illustrated by the flow chart in NEI 00-04, figure 5-6, Other External Hazards. NEI 00-04, figure 5-6 provides guidance to be used to determine the safety significance of SSCs for other external hazards (excluding internal fires and seismic hazards). The NRC staff finds that the licensee will assess the risk from all other external hazards consistent with figure 5-6 of NEI 00-04 as endorsed in RG 1.201, Revision 1.

Regarding tornado missile protection, the licensee confirmed in response to NRC staffs RAI that an evaluation determined that Waterford had no non-conforming or degraded barriers for SSCs. The NRC staff finds that the licensees evaluation is adequate for this application regarding the tornado missile hazard.

In summary, the use of the Waterford IPEEE results described by the licensee in the LAR, its RAI responses, and the licensees assessment of other external hazards (i.e., high winds, tornadoes, and external flood) is consistent with section 5 of NEI 00-04, Revision 0, as

endorsed in RG 1.201, Revision 1. Therefore, the NRC staff concludes that the licensees treatment of other external hazards is acceptable and meets the requirements of 10 CFR 50.69(c)(1)(ii).

3.3.1.3 Component Safety Significance Assessment for Passive Components In section 3.1.2, Passive Categorization Process, of the enclosure to the LAR, the licensee proposed using a categorization method, for passive components not cited in NEI 00-04, Revision 0, or RG 1.201, Revision 1, respectively, for passive component categorization, but was approved by the NRC for ANO-2. The ANO-2 methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference 26). The ANO-2 methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance is generally measured by the frequency and the consequence of, in this case, pipe ruptures. Treatment requirements (including repair/replacement) only affect the frequency of passive component failure. Categorizing passive components solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. Therefore, the NRC staff finds that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.

In section 3.1.2, Passive Categorization Process, of the enclosure to the LAR, the licensee stated, in part, [t]he passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. Consistent with ANO2-R&R-004, Class 1 pressure retaining SSCs in the scope of the system being categorized will be assigned HSS and cannot be changed by the IDP. The NRC staff finds the licensees proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

3.3.1.4 Key Principle 4 Conclusions Based on the NRC staffs review for IEPRA (includes internal floods) and FPRA acceptability and the evaluation of the use of non-PRA methods, set forth above, the NRC staff concludes that the proposed change satisfies the fourth key principle for risk informed decision making prescribed in RG 1.174, Revision 3.

3.3.2 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04, Revision 0, provides guidance that includes programmatic configuration control and a periodic review to ensure that the all aspects of the 10 CFR 50.69 program (i.e., includes traditional engineering analyses) and PRA models used to perform the risk assessment continue to reflect the as-built-as-operated plant, and that plant modifications and updates to the PRA overtime are continually incorporated.

Sections 11 and 12 of NEI 00-04, Revision 0, includes a discussion on periodic review; and program documentation and change control. Maintaining change control and periodic review will

also maintain confidence that all aspects of the 10 CFR 50.69 program and risk categorization for SSCs, continually reflect the Waterford 3 as-built, as-operated plant.

The NRC staff finds the risk management process described by the licensee in the LAR is consistent with section 12 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and consistent with the requirements in 10 CFR 50.69(e). Based on the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk informed decision making prescribed in RG 1.174, Revision 3.

4.0 PROPOSED REVISION TO THE OPERATING LICENSE The licensee proposed the following amendment to the RFOL for Waterford 3. The proposed License Condition 2.C.22 would state:

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergys submittal letter dated December 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 269 dated November 30, 2022.

Entergy will complete closure of the four Human Reliability Analysis (HRA) Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4 in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022, using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The NRC staff finds that the proposed license condition is acceptable, because: (1) it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that has previously been endorsed by the NRC; and (2) the evaluation

in SE Section 3.3.1.2, finds the non-PRA methods for assessing risk for seismic and passive components which are deviations from NEI 00-04, to be acceptable.

The NRC staff notes that the guidance for implementing 10 CFR 50.69 provided by the Commission in the Federal Register notice dated November 22, 2004,2 section III.4.10.2, Section 50.36 Technical Specifications, stated that the 10 CFR 50.69 rule does not include 10 CFR 50.36 in the list of special treatment requirements that may be replaced by the alternative 10 CFR 50.69 requirements for RISC-3 and RISC-4 SSCs when implementing a 10 CFR 50.69 license amendment. As a result, the NRC staff does not consider the Technical Specifications (including Improved Technical Specifications) and the associated Technical Requirements Manual to be part of the 10 CFR 50.69 rule. Therefore, the licensee must address proposed changes to its Technical Specifications separately.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Louisiana State official was notified of the proposed issuance of the amendment on November 10, 2022. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on March 23, 2021 (86 FR 15501), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1 Gaston, R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, dated 2 Federal Register Notice (69 FR 68008, 68028-68029; November 22, 2004), related to Risk-Informed Categorization and Treatment of Structure, Systems and Components for Nuclear Power Reactors.

December 18, 2020 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20353A433).

2 Gaston, R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Admendment Request to Adopt of 10 CFR 50.69, dated October 1, 2021 (ML21274A876).

3 Couture, P., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, Supplement to Application to Adopt of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, dated April 25, 2022 (ML22115A062).

4 Couture, P, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-505, dated August 19, 2022 (ML22231B160).

5 Couture, P., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information to License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk informed Extended Completion Times - RITSTF Initiative 4b, dated May 16, 2022 (ML22136A310).

6 Drake, J., U.S. Nuclear Regulatory Commission, email to R. Devoe, Entergy Operations, Inc., Final RAIs to Entergy Operations, Waterford Steam Electric Station, Unit 3 LAR to Adopt 10 CFR 50.69, dated July 26, 2021 (ML21218A040).

7 Drake, J., U.S. Nuclear Regulatory Commission, email to J. Lewis, Entergy Operations, Inc.,

Request for Additional Information: Waterford 3 - License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-505, dated July 22, 2022 (ML22206A017).

8 U.S. Nuclear Regulatory Commission, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Regulatory Guide 1.201, Revision 1, dated May 2006 (ML061090627).

9 U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, dated March 2009 (ML090410014).

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13 Nuclear Energy Institute, 10 CFR 50.69 SSC Categorization Guideline, NEI 00-04, Revision 0, dated July 2005 (ML052910035).

14 Electric Power Research Institute, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, EPRI Report 3002017583, dated February 2020 (ML21082A170).

15 Markley, M. T., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc.,

Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for

Repair/Replacement Activities in Class 2 and 3, Moderate and High Energy Systems, dated April 22, 2009 (ML090930246).

16 U.S. Nuclear Regulatory Commission, Individual Plant Examination of External Events (IPEEEs) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Generic Letter 88-20, Supplement 4), dated June 28, 1991 (ML031150485).

17 American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS),

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

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19 Anderson, V. K., Nuclear Energy Institute, letter to S. 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 (ML17086A431).

20 U.S. Nuclear Regulatory Commission and Electric Power Research Insitute, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009, NUREG-2169 and EPRI 3002002936, dated January 2015 (ML15016A069).

21 Reisi-Fard, M., U.S. Nuclear Regulatory Commission, memorandum to J. G. GItter, U.S.

Nuclear Regulatory Commission, 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, dated May 30, 2017 (ML17031A269).

22 Lawrence Livermore National Laboratory for U. S. Nuclear Regulatory Commission (NRC),

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23 Vaidya, B. K., U.S. Nuclear Regulatory Commission, letter to D. P. Rhoades, Exelon Generation Company, LLC, LaSalle County Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 249 and 235 Related to Application to Adopt 10 CRF 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated May 27, 2021 (ADAMS Accession No. ML21082A422).

24 Murray, D., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding LasSalle License Amendment Request to Renewed Facility Operatng License to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components, dated October 16, 2020 (ML20290A791).

25 Murray, D., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69, dated January 22, 2021 (ML21022A130).

26 American Society of Mechanical Engineers, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activitie, ASME Code Case N-660, July 2002.

Principal Contributors: M. Biro, A. Schwab D. Wu S. Park Date: November 30, 2022

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