ML24269A083

Issuance of Amendment Nos. 247 and 249 Adoption of 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML24269A083
Person / Time
Site:	Diablo Canyon
Issue date:	12/11/2024
From:	Samson Lee Plant Licensing Branch IV
To:	Gerfen P Pacific Gas & Electric Co
Lee S, 301-415-3158
References
EPID L-2023-LLA-0137
Download: ML24269A083 (1)

Text

December 11, 2024 Paula Gerfen Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424

SUBJECT:

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 247 AND 249 RE: ADOPTION OF 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2023-LLA-0137)

Dear Paula Gerfen:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 247 to Facility Operating License (FOL) No. DPR-80 and Amendment No. 249 to FOL No. DPR-82 for the Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon),

respectively. The amendments consist of changes to the FOLs in response to your application dated September 27, 2023, as supplemented by letter dated August 8, 2024.

The amendments revise Diablo Canyon FOL Nos. NPF-80 and NPF-82 to add a new license condition to allow for 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. The provisions in 10 CFR 50.69 allows adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated, systematic, risk-informed process for categorizing SSCs according to their safety significance.

A copy of the related safety evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 247 to DPR-80

2. Amendment No. 249 to DPR-82

3. Safety Evaluation cc: Listserv

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 247 License No. DPR-80

1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated September 27, 2023, as supplemented by letter dated August 8, 2024, 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.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Facility Operating License No. DPR-80 is hereby amended to add paragraph 2.C.12 to read as follows:

10 CFR 50.69 The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; 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 Individual Plant Examination of External Events (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.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 365 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony T. Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-80 Date of Issuance: December 11, 2024 Tony T.

Nakanishi Digitally signed by Tony T. Nakanishi Date: 2024.12.11 15:50:45 -05'00'

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 249 License No. DPR-82

1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated September 27, 2023, as supplemented by letter dated August 8, 2024, 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.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Facility Operating License No. DPR-82 is hereby amended to add paragraph 2.C.12 to read as follows:

10 CFR 50.69 The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; 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 Individual Plant Examination of External Events (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.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 365 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony T. Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-82 Date of Issuance: December 11, 2024 Tony T.

Nakanishi Digitally signed by Tony T. Nakanishi Date: 2024.12.11 15:51:15 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 247 TO FACILITY OPERATING LICENSE NO. DPR-80 AND LICENSE AMENDMENT NO. 249 TO FACILITY OPERATING LICENSE NO. DPR-82 DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of Facility Operating License Nos. DPR-80 and DPR-82 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Amendment No. 22, 120, 135, 141, 201, 225, 230, 247 PG&E shall keep the staff informed on the progress of the reevaluation program as necessary, but as a minimum will submit quarterly progress reports and arrange for semi-annual meetings with the staff. PG&E will also keep the ACRS informed on the progress of the reevaluation program as necessary, but not less frequently than once a year.

(8)

Control of Heavy Loads (SSER 27,Section IV.6)

Prior to startup following the first refueling outage, the licensee shall submit commitments necessary to implement changes and modifications as required to satisfy the guidelines of Section 5.1.2 through 5.1.6 of NUREG-0612 (Phase II: 9-month responses to the NRC Generic Letter dated December 22, 1980).

(9)

Emergency Preparedness (SSER 27,Section IV.3)

In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of preparedness, the provisions of 10 CFR Section 50.54( s)(2) will apply.

(10)

Masonry Walls (SSER-27,Section IV.4: Safety Evaluation of November 2, 1984)

Prior to start-up following the first refueling outage, the licensee shall (1) evaluate the differences in margins between the staff criteria as set forth in the Standard Review Plan and the criteria used by the licensee, and (2) provide justification acceptable to the staff for those cases where differences exist between the staffs and the licensee's criteria.

(11)

Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel pool as described in the application dated October 30, 1985 (LAR 85-13) as supplemented.

Amendment No. 8 issued on May 30, 1986 and stayed by the U.S. Court of Appeals for the Ninth Circuit pending completion of NRC hearings is hereby reinstated.

Prior to final conversion to the modified rack design, fuel may be stored, as needed, in either the modified storage racks described in Technical Specification 5.6.1.1 or in the unmodified storage racks (or both) which are designed and shall be maintained with a nominal 21-inch center-to-center distance between fuel assemblies placed in the storage racks.

(12) 10 CFR 50.69 The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; the shutdown safety assessment Amendment No. 225, 247 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 Individual Plant Examination of External Events (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.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

(13)

Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 230 are hereby incorporated into this license. Pacific Gas and Electric Company shall operate the facility in accordance with the Additional Conditions.

D.

Exemption Exemption from certain requirements of Appendix J to 10 CFR Part 50 is described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 9. This exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest. Therefore, this exemption, previously granted in Facility Operating License No. DPR-76, is hereby reaffirmed. The facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission.

E.

Physical Protection The licensee 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 combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: Diablo Canyon Power Plant, Units 1 and 2 Physical Security Plan, by Training and Qualification Plan, and Safeguards Contingency Plan, submitted by letter dated May 16, 2006.

PG&E 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 PG&E CSP was approved by License Amendment No. 210, as supplemented by a change approved by License Amendment No. 220.

Pursuant to NRCs Order EA-13-092, dated June 5, 2013, NRC reviewed and approved the license amendment 222 that permitted the security personnel of the licensee to possess and use certain specific firearms, ammunition, and other Amendment No. 82, 75, 97, 129, 188, 189, 193, 225, 247 Revised by letter dated July 11, 2007 devices, such as large-capacity ammunition feeding devices, notwithstanding local, State, and certain Federal firearms laws that may prohibit such possession and use.

F.

Deleted.

G.

Deleted.

H.

Financial Protection PG&E 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.

I.

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders Amendment No. 247 J.

Term of License This License is effective as of the date of Issuance and shall expire at midnight on November 2, 2024.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by:

Edson G. Case for Harold R. Denton, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A - Technical Specifications

2. Appendix B - Environmental Protection Plan

3. Appendix C - Deleted

4. Appendix D - Additional Conditions Date of Issuance: November 2, 1984

Amendment No. 227, 232, 249 (10) Pipeway Structure DE and DDE Analysis (SSER 32, Section 4)

Prior to start-up following the first refueling outage PG&E shall complete a confirmatory analysis for the pipeway structure to further demonstrate the adequacy of the pipeway structure for load combinations that include the design earthquake (DE) and double design earthquake (DDE).

(11)

Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel pool as described in the application dated October 30, 1985 (LAR 85-13) as supplemented.

Amendment No. 6 issued on May 30, 1986 and stayed by the U.S. Court of Appeals for the Ninth Circuit pending completion of NRC hearings is reinstated.

Prior to final conversion to the modified rack design, fuel may be stored, as needed, in either the modified storage racks described in Technical Specification 5.6.1.1 or in the unmodified storage racks (or both) which are designed and shall be maintained with a nominal 21-inch center-to-center distance between fuel assemblies placed in the storage racks.

(12) 10 CFR 50.69 The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; 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 Individual Plant Examination of External Events (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.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

(13)

Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 232, are hereby incorporated into this license. Pacific Gas and Electric Company shall operate the facility in accordance with the Additional Conditions.

Amendment No. 227, 249 D.

Exemption (SSER 31, Section 6.2.6)

An exemption from certain requirements of Appendix J to 10 CFR Part 50 is described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 9. This exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest. Therefore, this exemption previously granted in Facility Operating License No. DPR-81 pursuant to 10 CFR 50.12 is hereby reaffirmed. The facility will operate, with the exemption authorized, in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission.

E.

Physical Protection The licensee 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 provision 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 combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Diablo Canyon Power Plant, Units 1 and 2 Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan," submitted by letter dated May 16, 2006.

PG&E 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 PG&E CSP was approved by License Amendment No. 212, as supplemented by a change approved by License Amendment No. 222.

Pursuant to NRCs Order EA-13-092, dated June 5, 2013, NRC reviewed and approved the license amendment 224 that permitted the security personnel of the licensee to possess and use certain specific firearms, ammunition, and other devices, such as large-capacity ammunition feeding devices, notwithstanding local, State, and certain Federal firearms laws that may prohibit such possession and use.

F.

Deleted.

G.

Deleted.

H.

Financial Protection PG&E 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.

Amendment No. 96, 118, 190, 191, 194, 227, 249 Revised by letter dated December 11, 2006 Revised by letter dated July 11, 2007 I.

Mitigation Strategy Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders J.

Term of License This License is effective as of the date of issuance and shall expire at midnight on August 26, 2025.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by: Harold R. Denton Harold R. Denton, Director Office of Nuclear Reactor Regulation Attachments:

1.

Appendix A Technical Specifications (NUREG-1151)

2.

Appendix B - Environmental Protection Plan

3.

Appendix C - Deleted

4.

Appendix D - Additional Conditions Date of Issuance: August 26, 1985

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 249 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

AND PROPOSED LICENSE CONDITION By letter dated September 27, 2023 (Reference 1), as supplemented by letter dated August 8, 2024 (Reference 2), Pacific Gas and Electric Company (PG&E, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon). The amendments would allow the licensee to implement 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.

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

The licensee proposed the following license condition to the Facility Operating Licenses for Diablo Canyon:

The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; 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 Individual Plant Examination of External Events (IPEEE) 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 for other external hazards.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

In email correspondence dated July 2, 2024, the NRC staff requested additional information from the licensee (Reference 3). The licensee responded to the request for additional information (RAI) in a supplemental letter dated August 8, 2024 (Reference 2).

The supplemental letter dated August 8, 2024, 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 (FR) on November 28, 2023 (88 FR 83168).

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 will perform their design-basis functions. For SSCs with a function that is categorized as low safety significant (LSS),

alternative treatment requirements may be implemented in accordance with the regulation. For SSCs with a function determined to be high safety significant (HSS), requirements may not be reduced.

The regulation, 10 CFR 50.69, 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 process is employed to determine the safety significance of SSCs and assign each one to a particular RISC.

Categorization of an SSC does not alter its functional requirements. Equipment that is required to perform a design-basis function may not be removed from the facility. Instead, 10 CFR 50.69 enables 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 and may be 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 HSS equipment.

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

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 4)

RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 5)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 6)

RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference 7)

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference 8)

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP) chapter 19, section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (Reference 9) 2.3 NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 10), as endorsed by RG 1.201, Revision 1 (Reference 7),

for trial use with clarifications.1 The guideline describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process uses PRA and other methods to determine the safety significance of SSCs and categorizes them into one of four RISC categories defined in that regulation.

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

Sections 3.2, Use of Risk Information, and 5.1, Internal Events Assessment, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i), which requires that the process consider a plant-specific PRA.

Sections 3, Assembly of Plant-Specific Inputs; 4, System Engineering Assessment; 5, Component Safety Significance Assessment; and 7 Preliminary Engineering Categorization of Functions, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii), which requires an integrated, systematic process.

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

Section 8, Risk Sensitivity Study, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv), which requires maintaining sufficient safety margins and ensuring that increases in risk are small.

1 Regulatory Guide 1.201 (Reference 7) describes SSC categorization as an integrated decision-making process that incorporates both PRA and non-PRA evaluations of safety significance. Licensees must use risk evaluations and insights that cover the full spectrum of potential events and the entire range of plant operating modes.

Section 2, Overview of Categorization Process, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v), which requires that the process address entire systems and structures rather than selected components.

Sections 9, IDP [Integrated Decisionmaking Panel] Review and Approval; and 10,SSC Categorization, provide specific guidance corresponding to 10 CFR 50.69(c)(2), which requires an IDP with applicable expertise. 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),

Feedback and process adjustment. Maintaining change control and conducting periodic reviews provides confidence that all aspects of the program reasonably reflect both 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).

The Nuclear Management and Resources Council Inc. (NUMARC, a predecessor to the NEI) published NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 11). These guidelines are incorporated in NEI 00-04 (Reference 10) for use in the categorization process to address shutdown risk. The intent of this document was to provide guidance to utilities on assessing and enhancing practices for planning and conducting outages.

An important premise of the document is that a full understanding of vulnerabilities present during shutdown conditions will enhance safety during shutdown.

3.0 TECHNICAL EVALUATION

3.1 Categorization Process Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decision-making process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform. They are placed into one of the following four RISC categories:

RISC-1: Safety-related SSCs that perform safety significant functions2 RISC-2: Non-safety-related SSCs that perform safety significant functions RISC-3: Safety-related SSCs that perform low safety significant functions RISC-4: Non-safety-related SSCs that perform low safety significant functions The SSCs have functions that are HSS or LSS, and they are classified accordingly. For SSCs with HSS functions (i.e., RISC-1 and RISC-2 SSCs), 10 CFR 50.69 maintains current regulatory requirements for special treatment, that is, all existing special treatment requirements remain in force. For SSCs with LSS functions, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2). 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.

Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission 2 NEI 00-04 10, Revision 0, uses the term high-safety-significant to refer to SSCs that perform safety-significant functions. The NRC understands HSS to have the same meaning as safety-significant as used in 10 CFR 50.69, which applies to RISC-1 and RISC-2 SSCs.

determines that the categorization process satisfies the requirements of 10 CFR 50.69(c). As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for LSS SSCs:

(i) 10 CFR Part 21 (ii) a portion of 10 CFR 50.46a(b)

(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)

(v) specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)

(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix)

Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 (x) specified requirements for containment leakage testing (xi) specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100 that relate to qualification testing and specific engineering methods for SSC design The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04 (Reference 10), 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, used the framework provided in RG 1.174, Revision 3 (Reference 4), and RG 1.201, Revision 1 (Reference 7).

Section 2 of NEI 00-04 states that the categorization process includes the following eight primary steps:

1.

Assembly of Plant-Specific Inputs

2.

System Engineering Assessment

3.

Component Safety Significance Assessment

4.

Defense-in-Depth Assessment

5.

Preliminary Engineering Categorization of Functions

6.

Risk Sensitivity Study

7.

Integrated Decision-Making Panel Review and Approval

8.

SSC Categorization 3.2 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed changes, including both permanent and temporary changes, is to show that the proposed changes to the licensing basis meet the five key principles of risk-informed decision-making stated in section C of RG 1.174, Revision 3 (Reference 4):

Principle 1:

The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.

Principle 2:

The proposed licensing basis change is consistent with the defense-in-depth philosophy.

Principle 3:

The proposed licensing basis change maintains sufficient safety margins.

Principle 4:

When proposed licensing basis changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5:

The impact of the proposed licensing basis change should be monitored by using performance measures strategies.

The NRC staff evaluates the first three principles using traditional engineering methods. The last two are evaluated in light of the licensees PRA and other methods of assessing risk.

3.3 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles identified in RG 1.174, Revision 3 (Reference 4) and are pertinent to: (1) compliance with current regulations, (2) evaluation of defense in depth, and (3) evaluation of safety margins.

3.3.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations In section 3.1.1, Overall Categorization Process, of the enclosure to its LAR (Reference 1), the licensee stated that it will implement the risk-informed categorization process in accordance with NEI 00-04 (Reference 10), as endorsed by RG 1.201, Revision 1 (Reference 7). The licensee provided further discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the endorsed guidance of NEI 00-04.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, as well as the monitoring outlined in NEI 00-04 (Reference 10), will ensure that the SSC functions continue to be met, that any performance deficiencies will be identified, and that appropriate corrective actions will be taken. The NRC staff finds that the licensees SSC categorization program includes the appropriate steps/elements prescribed in NEI 00-04 to assure that SSCs are appropriately categorized, consistent with 10 CFR 50.69. Therefore, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision-making identified in RG 1.174, Revision 3.

3.3.2 Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth Philosophy In RG 1.174, Revision 3, the NRC staff identified the following considerations used for evaluating whether a proposed licensing basis change is consistent with the defense-in-depth 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.

In section 3.1.1 of the LAR enclosure (Reference 1), the licensee clarified that, consistent with the guidance in NEI 00-04 (Reference 10), it would require an SSC categorized as HSS based on the defense-in-depth assessment to be categorized HSS per the final categorization, and that cannot be changed by the IDP.

The NRC staff finds that the licensees process is consistent with the NRC-endorsed guidance in NEI 00-04 and concludes that the proposed change is consistent with the defense-in-depth philosophy. For this reason, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the second key principle for risk-informed decision-making identified in RG 1.174, Revision 3 and fulfills the requirement of 10 CFR 50.69(c)(1)(iii) that defense in depth be maintained.

3.3.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins Key risk-informed principle in RG 1.174 states that the licensing basis change should maintain sufficient safety margins. 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 SSCs to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, providing justification that sufficient safety margin will continue to exist.

The NRC staff notes that the design-basis function of SSCs as described in the plants licensing basis, including the Diablo Canyon Updated Final Safety Analysis Report and plant technical specification bases, do not change and the safety margins described 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 safety margins are maintained by the proposed methodology, and the third key safety principle identified in RG 1.174, Revision 3 (Reference 4) is satisfied.

In section 2.2, Reason for the Proposed Change, of the enclosure to its LAR (Reference 1),

the licensee stated, The safety functions [in the categorization process] include the design basis functions, as well as functions credited for severe accidents (including external events).

Section 3.1.1 of the LAR enclosure summarizes the different hazards and plant states for which functional and risk significant information will be collected. In the same section, the licensee confirmed that the SSC categorization process documentation will include, among other items, system functions, identified and categorized with the associated bases, and mapping of components to support functions.

The NRC staff finds that the process described in the LAR is consistent with NEI 00-04 as endorsed by the NRC in RG 1.201, Revision 1 (Reference 7). Therefore, it meets the requirements set forth in 10 CFR 50.69(c)(1)(ii) and 10 CFR 50.69(c)(1)(iv).

3.4 Risk-Informed Assessment 3.4.1 Probabilistic Risk Assessment 3.4.1.1 Scope of the PRA In section 3.2, Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii)), of the enclosure to its LAR (Reference 1), the licensee described the Diablo Canyon PRA, which comprises a full-power, level 1, internal events PRA (including internal floods), a fire PRA, and a seismic PRA.

Each PRA evaluates risk metrics of core damage frequency (CDF) and large early release frequency (LERF).

In attachment 3, Technical Acceptability of the Diablo Canyon Power Plant PRA Models, of the enclosure to its LAR, the licensee described the peer review of each PRA, as well as associated finding closure reviews. Open findings were reviewed and closed 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 (F&Os) (Appendix X) dated February 21, 2017 (Reference 12). The NRC staffs evaluations of those peer reviews and closure reviews are discussed in section 3.4.1.2, below.

Aspects considered by the NRC staff to evaluate the scope of the PRA include: (1) a process for peer review and independent assessment, (2) history of peer reviews and their results, (3) credit taken in the PRA for the diverse and flexible coping strategy (FLEX), and (4) assessment of assumptions and approximations.

The information provided in the LAR (Reference 1) and the supplemental letter (Reference 2) is sufficient to support the NRC staff review of the Diablo Canyon PRA and therefore the NRC staff finds that it meets the requirements of 10 CFR 50.69(b)(2)(iii).

3.4.1.2 Peer Review of the PRA 3.4.1.2.1 Peer Review History of the Internal Events PRA In attachment 3 of the enclosure to the LAR, the licensee stated that the internal events PRA model (which includes internal floods) was subjected to a full-scope peer review in December 2012, consistent with RG 1.200, Revision 2 (Reference 6). In June 2023, the licensee conducted an independent assessment for closure of the finding-level F&Os. Several F&Os were determined to constitute upgrades, so the peer review team conducted a focused-scope review and concluded that the relevant supporting requirements were met. All finding-level F&Os were assessed in accordance with Appendix X (Reference 12). The independent assessment concluded that all finding-level F&Os for the internal events PRA have been closed and that no new methods were identified.

The NRC staff reviewed the results of this assessment and found that no new methods or upgrades were inadvertently incorporated into the internal events PRA without a peer review.

The NRC staff found that all F&Os were appropriately assessed by the independent

assessment team in accordance with 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 (the 2009 PRA standard)

(Reference 13) as endorsed in RG 1.200, Revision 2. Therefore, the NRC staff concludes that the Diablo Canyon internal events PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.1.2.2 Peer Review History of the Fire PRA The licensees fire PRA was subject to two peer reviews. The first was in January 2008 as part of the pilot application of the fire PRA peer review process of NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines (Reference 14). The 2008 peer review was conducted against the requirements of the ANS Standard ANSI/ANS-58.23-2007, FPRA Methodology (Reference 15), which was in effect at the time. A second peer review was completed in December 2010. This was conducted against the 2009 PRA standard (Reference 13), and it addressed elements that did not meet Capability Category II during the previous review.

In August and September of 2018, an independent assessment team reviewed all finding-level F&Os against the 2009 PRA standard, using Appendix X (Reference 12). Two F&Os were identified by PG&E as upgrades, and a focused-scope peer review was therefore conducted in conjunction with the closure review. The team determined that no other F&Os constituted an upgrade, and they did not identify the use of any new methods.

The NRC staff evaluated the reports of the peer review and independent assessment teams.

The staff found that they were consistent with the 2009 PRA standard (Reference 13) as endorsed in RG 1.200, Revision 2 (Reference 6), and they are therefore acceptable.

In section 3.2.2, Internal Fire Events, of the enclosure to its LAR, the licensee stated that The internal fire PRA model has been developed consistent with NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities (Reference 16) and uses only NRC-approved methods.

The NRC staff has reviewed the fire PRA peer review results and the licensees resolution of the results concludes that the licensees fire PRA was appropriately peer-reviewed, consistent with the 2009 PRA Standard (Reference 13) as endorsed in RG 1.200, Revision 2 (Reference 6),

and that the F&Os have been dispositioned appropriately. Therefore, the licensees fire PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.1.2.3 Peer Review History of the Seismic PRA In attachment 3 of the enclosure to the LAR, the licensee stated that its seismic PRA model received a full-scope peer review in September 2017. The licensee explained that the peer review included a review of the seismic hazard and fragility analyses and was performed consistent with RG 1.200, Revision 2 (Reference 6), using the PRA Standard ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 17) (the 2013 PRA Standard, identified as Addendum B in the LAR).

The licensee stated that an independent assessment of the finding-level F&Os was conducted in October through December 2017. The licensee further stated that three F&Os were identified by the licensee as upgrades and two additional F&Os were identified by the assessment team as upgrades. Consequently, a focused-scope peer review was conducted in conjunction with the closure review. The licensee concluded that (a) all applicable supporting requirements of Addendum B were met and (b) that at the conclusion of the independent assessment and focused-scope peer review, supporting requirements satisfied at least Capability Category II.

In RG 1.200, Revision 2, the NRC staff endorsed the 2009 PRA Standard (Reference 13).

However, the staff has not endorsed the 2013 PRA Standard (Reference 17). During its review of the licensees LAR to revise technical specifications to adopt risk-informed completion times (Reference 18), the staff requested that the licensee provide a comparison of the supporting requirements in the 2013 PRA Standard with the supporting requirements in the 2009 PRA Standard to demonstrate that the supporting requirements of the 2009 PRA Standard have been met for instances where the criteria between the two standards differ.

In the licensees supplement to the LAR to revise technical specifications to adopt risk-informed completion times (Reference 19), the licensee compared the supporting requirements in the 2013 PRA Standard (Reference 17) with the supporting requirements in the 2009 PRA Standard (Reference 13). In the RAI, in support of the present LAR (Reference 3), the NRC staff requested that the licensee confirm that its comparison of the supporting requirements in the 2013 PRA Standard with the supporting requirements in the 2009 PRA Standard is valid for this LAR as well. In its RAI response (Reference 2), the licensee confirmed that its comparison of the supporting requirements in the 2013 PRA Standard with the supporting requirements in the 2009 PRA Standard is valid for the present LAR. The NRC staffs review and evaluation of this comparison is included in the safety evaluation for the license amendment approving the licensees request to revise technical specifications to adopt risk-informed completion times (Reference 20).

Based on its review (Reference 20) of the licensees comparison of the supporting requirements in the 2013 PRA Standard (Reference 17) with the supporting requirements in the 2009 PRA Standard (Reference 13), the NRC staff finds that the use of the 2013 PRA Standard is an acceptable alternative to the NRC-endorsed approach for the licensees seismic PRA used to support this application, all finding-level F&Os were closed, and the licensees evaluation of the key assumptions and sources of uncertainty for its seismic PRA is consistent with RG 1.200, Revision 2 (Reference 6). Therefore, the licensees seismic PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.1.3 Credit for Diverse and Flexible Coping Strategy (FLEX) Equipment The NRC staff has identified challenges to incorporating FLEX equipment and strategies into a PRA model to support risk-informed decision-making. The NRC staff assessment of industry guidance for crediting mitigation strategies in PRA is documented in a memorandum dated May 6, 2022, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments (Reference 21).

The licensee stated in its LAR that its PRAs models do not credit risk reduction through use of FLEX equipment, but also stated that the PRA models operator actions for two FLEX strategies that are addressed in the seismic PRA. These actions include credit for FLEX strategies to monitor steam generator level at the hot shutdown panel without instrument alternating current

power available during a station black out (whether induced by a seismic event or with loss of all direct current power).

The NRC staff finds that the licensees internal events PRA (includes internal floods), fire PRA, and seismic PRA do not credit FLEX equipment for the SSC categorization process. Although FLEX strategies were modeled, no new methods or upgrades were inadvertently incorporated into the PRA without a peer review. This is consistent with the NRC-endorsed 2009 PRA standard (Reference 13) and is therefore acceptable for risk-informed categorization of SSCs.

3.4.1.4 Assessment of Assumptions and Approximations 3.4.1.4.1 Identification of Key Assumptions and Sources of Uncertainty In attachment 5, Disposition of Key Assumptions/Sources of Uncertainty, of the enclosure to its LAR (Reference 1), the licensee stated that NUREG-1855 (Reference 8) and Electric Power Research Institute (EPRI) 1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments (Reference 22) were used to identify, screen, and characterize sources of model uncertainty. Substep E-1.4 of NUREG-1855 is a qualitative screening process that involves identifying those sources of uncertainty and related assumptions in the base PRA that are relevant to this application. The licensee followed section 3.1.1 of EPRI 1016737 to evaluate identified model uncertainties using consensus models.3 The NRC staff finds that the assessment performed to identify the key assumptions/sources of uncertainty is consistent with the guidance in NUREG-1855 and EPRI 1016737, and it is therefore acceptable.

3.4.1.4.2 Treatment of the Key Assumptions and Sources of Uncertainty NUREG-1855 (Reference 8) provides guidance on how to address PRA uncertainties to assure that the risk-informed decision is in the context of the application for the decision under consideration. The licensee confirmed that sensitivity studies will be performed during 50.69 categorization consistent with table 5-2, Sensitivity Studies for Internal Events PRA, of NEI 00-04 (Reference 10). In accordance with that guidance, the results of the sensitivity study are to be given to the IDP for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

The licensee will perform routine PRA changes and updates to assure the PRA continually reflects the as-built, as-operated plant, in addition to changes made to the PRA to support the context of the analysis being performed (i.e., sensitivities). Paragraphs 50.69(e) and (f) of 10 CFR stipulate the process for feedback and adjustment to assure configuration control is maintained for these routine changes and updates to the PRA. The NRC staff evaluation of this activity is documented in section 3.4.4, below.

In attachment 5 to the LAR enclosure (Reference 1), the licensee supplied a table describing the disposition of key assumptions and sources of uncertainty that affect risk-informed categorization of SSCs. How these assumptions and sources of uncertainty had been identified is described in section 3.4.1.4.1 above. For some of these, the licensee found that the impact was negligible; for others, the licensee proposed sensitivity studies to confirm that appropriate 3 As defined in NUREG-1855, a consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group.

treatment would be applied. The NRC staff requested additional information to confirm that the dispositions are acceptable.

In its RAI (Reference 3), the NRC staff asked the licensee to describe how the interdependence of structures and systems of the opposite unit will be addressed when conducting sensitivity analysis of shared components.

In its response to the RAI (Reference 2), the licensee stated that implementation of 10 CFR 50.69 will address the interdependence of structures and systems that affect both units. The model that will be used for categorization results will include cross-unit impacts.

In its RAI (Reference 3), the NRC staff asked the licensee to clarify the modeling and success criteria used for high-pressure pumps in medium loss-of-coolant accidents, to justify the method used for modeling the case where two of two charging pumps might be required, and to explain why this approach does not affect the categorization results.

In its response to the RAI (Reference 2), the licensee explained how its modeling approach led to the use of conservative success criteria: one of two charging pumps and one of two safety injection pumps. The licensee also explained that this conservatism was addressed by applying an appropriate recovery factor to approximate the reliability of two charging pumps with safety injection not available. Although the licensee evaluated this approach as unlikely to have an impact on the categorization results, the licensee stated an assessment will be performed as part of each system categorization.

In its RAI (Reference 3), the NRC staff asked the licensee to justify the conclusion that the reduction of mission time for the diesel generators does not have a significant impact on the baseline PRA and categorization results. In its response to the RAI (Reference 2), the licensee explained that a convolution method electric power recovery model was used to quantify the probability distribution for the cumulative onsite power failure probability at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Onsite power failure and electric power nonrecovery probability were integrated over the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the loss of offsite power; therefore, the full 24-hour mission time is accounted for in this approach.

In its RAI (Reference 3), the NRC staff asked the licensee to justify the expectation that auxiliary salt water vacuum relief valves (vacuum breakers) cannot fail with an adverse impact on the auxiliary salt water system function. In its response to the RAI (Reference 2), the licensee explained that the valves are highly reliable mechanical components with a demonstrated failure probability far lower than other components in the same train. The licensee concluded that the vacuum breakers have a negligible impact on train failure and system categorization.

In its RAI (Reference 3), the NRC staff asked the licensee to identify those systems that are assumed to be failed in the fire and seismic PRA. Moreover, the licensee was asked to explain how the effect of this conservatism on categorization will be addressed. In its response to the RAI (Reference 2), the licensee offered an example of how this might be done but also stated that it will assess the impact of the assumption for each system categorized under 10 CFR 50.69, such as performing sensitivity studies. The results of this assessment are to be considered by its IDP as part of the categorization process.

In each case where assumptions and sources of uncertainty had been determined to have negligible impact, the NRC staff concluded that the licensee had an acceptable basis for determining that categorization of SSCs would not be affected. In all other cases, a sensitivity

study would be used to inform decision-making. The sensitivity studies will be consistent with table 5-2 of the guidance in NEI 00-04 (Reference 10) and the NRC staff finds that they will therefore be acceptable for categorization of SSCs.

3.4.1.5 PRA Importance Measures and Integrated Importance Measures The scope of modeled hazards for Diablo Canyon includes the internal events PRA (includes internal floods), fire PRA, and seismic PRA; and 10 CFR 50.69(c)(1)(ii) requires that the SSC functional importance be determined using an integrated, systematic process. Section 5.6, Integral Assessment, of NEI 00-04 (Reference 10), discusses the need for an integrated computation using available importance measures. It further states that the integrated importance measure essentially weights the importance from each risk contributor (e.g., internal events, fire, seismic PRAs) by the fraction of the total core damage frequency [or large early release frequency] contributed by that contributor. The guidance provides formulas to compute the integrated Fussell-Vesely importance (FV) and integrated risk achievement worth (RAW).

In its RAI (Reference 3), the NRC staff asked the licensee to explain how the integration of importance measures across hazards for the 10 CFR 50.69 categorization process will be performed.

In its response to the RAI (Reference 2), the licensee stated that the process to categorize each system will be consistent with the guidance in NEI 00-04. The approach described under the heading Integrated Importance Assessment in section 1.5 of NEI 00-04, states, In order to facilitate an overall assessment of the risk significance of SSCs, an integrated computation is performed using the available importance measures. This integrated importance measure essentially creates a weighted-average importance based on the importance measures and the risk contributed by each hazard (e.g., internal events, fire, seismic PRAs).

In its RAI (Reference 3), the NRC staff also asked the licensee to explain how the importance measures for the PRA models (e.g., FV and RAW) are derived and justify why the importance measures generated do not deviate from the NEI guidance or Table 3-1 of the LAR. The licensee was asked to provide justification to support why the integrated importance measures computed are appropriate for use in the categorization process if the practice or method used to generate the integrated importance measures deviates from the NEI guidance, In its response to the RAI (Reference 2), the licensee offered the following statement:

The process used to derive the importance measures for various hazards as described in Table 3-1 of [the LAR] is performed on a component basis in accordance with guidance in NEI 00-04 Section 5, Component Safety Significance Assessment. Some components in the Internal Events PRA may not be modeled explicitly in the Seismic PRA due to it being subsumed in a super component (as described in NEI 00-04), or because it was screened out of the analysis. For those that were screened out, the importance measure would be taken as non-risk significant (0 for FV or 1 for RAW). For the components that are included under a super component, the super component risk importance would instead be considered to determine risk significance. This treatment for components included under a super component is consistent with the NEI 00-04 Section 5, sentence As outlined in Section 1, by focusing on the significance of system functions and then correlating those functions to specific components

that support the function, it is possible to address even implicitly modeled components.

Finally, in its RAI (Reference 3), the NRC staff asked the licensee to explain how the importance measures for the seismic PRA (e.g., FV and RAW) are derived and to discuss how discretized basic events are combined to develop representative importance measures and compared to the importance measure thresholds in NEI 00-04 (Reference 10).

In its response to the RAI (Reference 2), the licensee stated that it will take each of the importance measures and create a weighted average based on its CDF/LERF weighting as described in NEI 00-04 (Reference 10), specifically section 1.5 Integrated Importance Assessment and section 5.6 Integral Assessment.

The NRC staff finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04 (Reference 10) as endorsed in RG 1.201, Revision 1 (Reference 7). Therefore, the NRC staff concludes that the licensees approach meets 10 CFR 50.69(c)(1)(ii).

3.4.1.6 PRA Acceptability Conclusions Pursuant to 10 CFR 50.69(c)(1)(i), the categorization process must consider results and insights from a plant-specific PRA. The use of the internal events PRA (includes internal floods), fire PRA, and seismic PRA to support SSC categorization is endorsed by RG 1.201, Revision 1 (Reference 7).

The PRAs must be acceptable to support the categorization process and must be subjected to a peer review process. The NRC staff evaluation of the peer reviews for Diablo Canyon is discussed in section 3.4.1.2, above.

The NRC staff finds the licensee provided sufficient information required for the NRC staff to conclude that the licensees PRA, comprising an internal events PRA, a fire PRA, and a seismic PRA, is acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201, Revision 1 (Reference 7). The key assumptions for the PRAs have been identified, consistent with the guidance in RG 1.200, Revision 2 (Reference 6) and NUREG-1855 (Reference 8), and addressed appropriately for this application. Therefore, the NRC staff concludes that the licensees PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and (ii).

3.4.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization The licensees categorization process uses the following non-PRA methods:

screening analysis performed for the IPEEE for high winds, external floods, and other hazards, updated to reflect part 6 of Addendum B (Reference 17) safe shutdown risk management program consistent with NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 11) risk-informed safety classification and treatment for repair/replacement activities in Class 2 and 3 moderate-and high-energy systems (ANO-2 passive categorization)

(Reference 23)

The NRC staffs review of these methods is discussed below.

3.4.2.1 Other 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 section 3.2.4 of the enclosure to the LAR (Reference 1), the licensee stated that it performed a systematic reevaluation of selected external hazards under the IPEEE and that it reexamined the external hazards in 2016 using the guidance in part 6 of the 2013 PRA Standard (Reference 17), to ensure the IPEEE conclusions remained bounding and to account for updated information. The licensee further stated that the results of the 2016 reexamination concluded that the external hazards (other than seismic) can be screened out; therefore, the there was no need for further detailed PRA of these external hazards.

In attachment 4, External Hazards Screening, of the enclosure to the LAR (Reference 1), the licensee provided a summary of its progressive screening approach applied for external hazards and a summary of the external hazards screening results. The hazards assessed in the LAR are those identified for consideration in nonmandatory Appendix 6-A of the 2009 PRA Standard (Reference 13).

In a supplement to the licensees LAR to revise technical specifications to adopt risk-informed completion times (Reference 19), the licensee provided additional information including the assumptions, data sources, methodology, and results for its assessment of the aircraft impact, extreme wind and tornado, hurricane, and tsunami external hazards. For each of these external hazards, the licensee performed a conservative analysis demonstrating the CDF has a mean frequency of less than 1x10-6 per year.

The NRC staff reviewed the licensees progressive screening approach for external hazards described in the enclosure to the LAR (Reference 1) and the additional information provided in the supplement to the licensees LAR to revise technical specifications to adopt risk-informed completion times (Reference 19). The NRC staff notes that the progressive screening criteria used in attachment 4 of the enclosure to the LAR are the same criteria presented in supporting requirements for screening external hazards EXT-B1 and EXT-C1 of the NRC-endorsed 2009 PRA Standard (Reference 13). Based on its review, the NRC staff concludes that the licensees treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

3.4.2.2 Shutdown Consistent with the guidance in NEI 00-04 (Reference 10), the licensee proposed using a shutdown safety assessment based on NUMARC 91-06 (Reference 11). This industry guidance addresses considerations for maintaining defense-in-depth for the five key safety functions during shutdown, namely, decay heat removal capability, inventory control, power availability, reactivity control, and containment. The guidance also specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and an alternative system/train to accomplish the given key safety function.

The use of NUMARC 91-06 (Reference 11) described by the licensee in the submittal is consistent with the guidance in NEI 00-04 (Reference 10) as endorsed by the NRC in RG 1.201, Revision 1 (Reference 7). The approach uses an integrated and systematic process to identify

HSS components. Therefore, the NRC staff finds that the licensees use of NUMARC 91-06 is acceptable, and meets the requirements set forth in 10 CFR 50.69(c)(1)(ii).

3.4.2.3 Component Safety Significance Assessment for Passive Components Passive components are not modeled in the PRA; therefore, a different assessment method is necessary to assess the safety significance of these components. Passive components are those components having only a pressure retaining function. This process also addresses the passive function of active components such as the pressure/liquid retention of the body of a motor-operated valve.

In section 3.1.2, Passive Categorization Process, of the enclosure to its LAR (Reference 1),

the licensee proposed using a categorization method for passive components which was not cited in NEI 00-04 (Reference 10) or RG 1.201, Revision 1 (Reference 7). This method was approved by the NRC for ANO-2. The licensee will evaluate passive components and the passive function of active components using the ANO-2 methodology for risk-informed repair/replacement activities, consistent with the related safety evaluation (Reference 23).

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). It uses a modification of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference 24).

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 a particular SSC or event, in this case, a pipe rupture.

However, treatment requirements (including repair and replacement) only affect the frequency of passive component failure. Categorizing the significance of pipe failure on the consequence of failure alone is conservative compared to the more general measure (which includes rupture frequency). The categorization will not be affected by changes in frequency arising from changes to the treatment.

In section 3.1.2 of the LAR enclosure (Reference 1), the licensee stated, The passive categorization process is intended to apply the same risk-informed process authorized by the NRC for the passive categorization of class 2, 3, and non-class components at ANO-2.... All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned HSS for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP.

That is, the proposed categorization methodology and ANO-2 precedent repair/replacement methodology do not allow an ASME Code Class 1 pressure retaining SSC (and supports, etc.)

to be recategorized from HSS to LSS. The NRC staff finds that the use of the repair/replacement methodology is consistent with the methodology evaluated and accepted previously in the case of ANO-2, as documented in the safety evaluation for acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs (Reference 23).

Accordingly, the NRC staff finds the licensees proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

3.4.3 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04 (Reference 10) address the fourth and fifth key principles of risk-informed decision-making identified in RG 1.174, Revision 3 (Reference 4), which pertain to the assessment of change in risk and monitoring the impact of the licensing basis change.

The NRC staff acknowledges that elements of the categorization process are not always performed in chronological order and may be performed in parallel. The licensees SSC categorization process is consistent with the guidance and methodology prescribed in NEI 00-04 (Reference 10) and endorsed in RG 1.201, Revision 1 (Reference 7), as summarized in the sections below.

3.4.3.1 Assembly of Plant-Specific Inputs The licensees risk categorization process uses PRAs to assess risks from internal events (including internal flooding), internal fire, and seismic hazards. Non-PRA methods are used to assess shutdown safety, passive component risk, and the significance of external hazards. For those that depart from the methodology prescribed in NEI 00-04, additional NRC staff review is discussed in section 3.4.2.3 of this safety evaluation.

The process used by the licensee is described in the LAR (Reference 1), as supplemented (Reference 2). The process for collecting and organizing information at the system level for defining boundaries, functions, and components is consistent with NEI 00-04 (Reference 10).

Because the process is consistent with NEI 00-04 as clarified and endorsed in RG 1.201, Revision 1 (Reference 7), the NRC staff finds that the process meets the requirements set forth in 10 CFR 50.69(c)(1)(v).

3.4.3.2 Component Safety Significance Assessment This step in the licensees categorization process assesses the safety significance of components. The licensee may use quantitative or qualitative risk information from a hazard addressed in a PRA model. Hazards that are not modeled with PRA can be assessed with other, non-PRA methods (e.g., seismic margin analysis). Alternatively, if they are determined to be negligible, they may be screened from further consideration. In the guidance of NEI 00-04 (Reference 10), component risk significance is assessed separately for the following hazard groups:

internal events (including internal floods) internal fire events seismic events external hazards (e.g., high winds, external floods) other hazards shutdown events passive categorization In section 3.1, Categorization Process Description (10 CFR 50.69(b)(2)(i)), of the LAR enclosure (Reference 1), the licensee described how its categorization process uses PRA to assess risks from internal events (including internal flood), internal fires, and seismic hazards.

For the other risk contributors, the licensees process uses the following non-PRA methods to characterize the risk:

Other external hazards were systematically evaluated under the IPEEE and reevaluated in 2016 with a preliminary screening and a quantitative screening using the criteria of the 2009 PRA Standard (Reference 13) as endorsed by the NRC in RG 1.200, Revision 2 (Reference 6).

Shutdown events are evaluated using a Safe Shutdown Risk Management program that is consistent with NUMARC 91-06 (Reference 11).

Passive Components: ANO-2 passive categorization methodology as approved by the NRC (Reference 23).

The approaches and methods proposed by the licensee to address internal events (including flooding), internal fires, seismic hazards, other external events, and shutdown events are consistent with the approaches and methods included in the guidance in NEI 00-04 (Reference 10). The NRC staffs evaluation of the seismic PRA is described in section 3.4.1.2.3 of this safety evaluation. The non-PRA method for the categorization for passive components is consistent with the ANO-2 methodology for passive components (Reference 23), which is 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 discussed in section 3.4.2.3 of this safety evaluation.

3.4.3.3 Risk Sensitivity Study (NEI 00-04, Section 8)

In section 3.1.1 of the enclosure to its LAR (Reference 1), the licensee acknowledges the statement in RG 1.201, Revision 1 (Reference 7) that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 (Reference 10)) must be followed to achieve reasonable confidence in the evaluations required by [10 CFR] 50.69(c)(1)(iv). The LAR goes on to state that an unreliability factor of 3 will be used for the studies described in NEI 00-04, specifically section 8, Risk Sensitivity Study.

In section 3.2.7, PRA Uncertainty Evaluations, of the enclosure to its LAR, the licensee further confirms that a cumulative sensitivity study will be performed where the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of 3.

This sensitivity study, together with the periodic review process, assure that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The NRC staff finds that the licensee will perform the risk sensitivity study consistent with the guidance in section 8 of NEI 00-04 and, therefore, will assure that the potential cumulative risk increase from the categorization is maintained acceptably low, as required by 10 CFR 50.69(c)(1)(iv).

3.4.3.4 Integrated Decision-Making NUREG-0800, Appendix B of SRP section 19.2 (Reference 9) provides guidance and the NRC staff expectations for the licensees integrated decision-making process. The appendix states in part, Risk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. NEI 00-04 guidance (Reference 10) identifies two steps in the categorization process: (1) Preliminary Engineering Categorization of Function, and (2) IDP Review and Approval. These call for the integrated assessment of the traditional engineering analyses and the risk results. Risk results come from the PRA and non-PRA assessments. Integrated decision-making is performed to make a determination of the safety significance of SSCs for their categorization. The staff reviews the two steps to ensure the processes is well-defined, systematic, repeatable, and scrutable.

3.4.3.5 Preliminary Engineering Categorization of Function (NEI 00-04, Section 7)

All the information collected and evaluated in the licensees engineering evaluations is provided to the IDP as described in section 7 of NEI 00-04 (Reference 10). The IDP will make the final decision about the safety significance of SSCs based on guidelines in NEI 00-04, the information they receive, and their expertise.

In section 3.1.1 of the enclosure to its LAR (Reference 1), the licensee acknowledged the NRC staffs clarification of NEI 00-04 that if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS.

The licensee also stated, Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS function components to LSS.

The NRC staff finds that the above description provided by the licensee for the preliminary categorization of functions is consistent with NEI 00-04 and is therefore acceptable.

3.4.3.6 IDP Review and Approval (NEI 00-04, Sections 9 and 10)

In section 3.1.1 of the enclosure to its LAR (Reference 1), the licensee stated that The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design engineering (mechanical and electrical), system engineering, safety analysis, and probabilistic risk assessment. Therefore, the IDP will comprise the expertise required by 10 CFR 50.69(c)(2).

The guidance in NEI 00-04 (Reference 10) provides confidence that the IDP expertise is sufficient to perform the categorization and that the results of the different evaluations (PRA and non-PRA) are used in an integrated, systematic process as required by 10 CFR 50.69(c)(1)(ii).

In section 3.1.1 of its LAR enclosure (Reference 1), the licensee affirms that At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA. In the LAR, the licensee also identifies how the IDP will be trained:

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, (1) the

purpose of the categorization; (2) the present treatment requirements for SSCs including requirements for design basis events; (3) PRA fundamentals; (4) the details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and (5) the defense-in-depth philosophy and requirements to maintain this philosophy.

The NRC staff finds that the licensees IDP areas of expertise meet the requirements in 10 CFR 50.69(c)(2) and the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04 (Reference 10).

3.4.3.7 Conclusion for Key Principle 4 The NRC staff reviewed the acceptability of the licensees internal events PRA (including internal floods), fire PRA, and seismic PRA. The staff also reviewed the use of PRA importance measures and integrated importance measures, the use of non-PRA methods, risk sensitivity studies, and integrated decision-making. Based on these reviews and the findings described in sections 3.4.3.1-6 above, the NRC staff has determined that the proposed change satisfies the fourth key principle for risk-informed decision-making identified in RG 1.174, Revision 3 (Reference 4) and ensures that any potential increases in risk are small as required by 10 CFR 50.69(c)(1)(iv).

3.4.4 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04 (Reference 10) provides guidance that includes programmatic configuration control and a periodic review to ensure that the all aspects of the 10 CFR 50.69 program (including 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 are continually incorporated.

Sections 11 and 12 of NEI 00-04 (Reference 10) include a discussion on periodic review, program documentation, and change control. The licensee commits to maintaining change control and periodic review, which will also maintain confidence that the risk categorization of SSCs will continue to reflect the as-built, as-operated plant.

3.4.4.1 Periodic Review Paragraph 50.69(e) of 10 CFR requires that periodic updates to the licensees PRA and SSC categorization must be performed. Changes over time to the PRA and to the SSC reliabilities are inevitable, and such changes are recognized by the 10 CFR 50.69(e) requirement for periodic updates.

In section 3.5, Feedback and Adjustment Process, of the enclosure to its LAR (Reference 1),

the licensee described the process for maintaining and updating the Diablo Canyon PRA models used for the 10 CFR 50.69 categorization process. Consistent with NEI 00-04, the licensee confirmed that the Diablo Canyon risk management process ensures that the PRA models used in this application continue to reflect the as-built and as-operated plant. The licensees process includes provisions for: monitoring issues affecting the PRA models (e.g.,

due to changes in the plant, errors or limitations identified in the model, and industry operational experience); assessing the risk impact of unincorporated changes; and controlling the model and associated computer files. The process also includes reevaluating previously categorized

systems to ensure the continued validity of the categorization. Routine PRA updates are performed every two refueling cycles at a minimum.

The NRC staff finds the risk management process described by the licensee in the LAR is consistent with the guidance in section 12 of NEI 00-04 (Reference 10). Considering the above, the staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision-making identified in RG 1.174 (Reference 4).

3.4.4.2 Program Documentation and Change Control Paragraph 50.69(f) of 10 CFR requires program documentation, change control, and records. In section 3.2.6, PRA Maintenance and Updates, of the enclosure to its LAR (Reference 1), the licensee stated that it will implement a process that addresses the requirements in section 11 of NEI 00-04 (Reference 10), pertaining to program documentation and change control records. In section 3.1.1 of the LAR enclosure, the licensee states that the RISC categorization process documentation will include the following 10 elements:

Program procedures used in the categorization System functions, identified and categorized with the associated bases Mapping of components to support function(s)

PRA model results, including sensitivity studies Hazards analyses, as applicable Passive categorization results and bases Categorization results including all associated bases and RISC classifications Component critical attributes for HSS SSCs Results of periodic reviews and SSC performance evaluations IDP meeting minutes and qualification/training records for the IDP members The NRC staff also recognizes that for facilities licensed under 10 CFR Part 50, Appendix B Criterion VI, Document Control, procedures are considered formal plant documents requiring that [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality.

The NRC staff finds that the elements provided in section 3.1.1 of the LAR enclosure for the Diablo Canyon 10 CFR 50.69 categorization process will be documented in formal licensee procedures consistent with section 11 of NEI 00-04, as endorsed by the NRC in RG 1.201, Revision 1 (Reference 7), and therefore will be sufficient for meeting the 10 CFR 50.69(f) requirement for program documentation, change control and records.

4.0 CHANGES TO THE OPERATING LICENSE Based on the NRC staffs review of the LAR and the licensees responses to the staffs RAIs, the staff identified specific actions, as described below, that are necessary to support the NRC staffs conclusion that the proposed program meets the requirements in 10 CFR 50.69 and the guidance in NEI 00-04 (Reference 10) as endorsed in RG 1.201, Revision 1 (Reference 7).

Note: Additional actions (e.g., final procedures and proposed alternative treatment) need not, and have not been developed, submitted, or reviewed by the NRC staff for issuance of the safety evaluation, but will be completed before implementation of the program as specified in 10 CFR 50.69(f)(4).

The NRC staffs finding on the acceptability of the PRA evaluation in the licensees proposed 10 CFR 50.69 process is conditioned upon the license condition provided below, which delineates completion of prerequisites identified in attachment 1 to the LAR (Reference 1). For the clarifications to the guidance of NEI 00-04 (Reference 10) and other changes that were described by the licensee, the NRC staff finds these to be routine and systematically addressed through the configuration management and control and periodic update processes as described in section 3.4 of this safety evaluation.

The licensee proposed the following license condition to the Facility Operating Licenses for Diablo Canyon, Units 1 and 2. The proposed 10 CFR 50.69 license condition states:

The Pacific Gas and Electric Company 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, internal fire, and seismic hazards; 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 Individual Plant Examination of External Events (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.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

The NRC staff finds that the proposed license condition is acceptable because it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable NRC and NRC-endorsed guidance.

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 (69 FR 68008, 68028-68029),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 (TSs) (including Improved Technical Specifications (ITS) and the associated Technical Requirements Manual to be part of the 10 CFR 50.69 rule. Therefore, the licensee needs to address proposed changes to its TS separately.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments on August 29, 2024. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, as published in Federal Register on November 28, 2023 (88 FR 83168), and there has been no public comment on such finding. Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1.

Petersen, D. B., Pacific Gas and Electric Company, letter to the U.S. Nuclear Regulatory Commission, License Amendment Request 23-02, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated September 27, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23270B909).

2.

Rogers, J. E., Pacific Gas and Electric Company, letter to the U.S. Nuclear Regulatory Commission, Response to Request for Additional Information for License Amendment Request 23-02, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated August 8, 2024 (ML24221A140).

3.

Lee, S, U.S. Nuclear Regulatory Commission, email to K. Schrader, Pacific Gas and Electric Company, Request for Additional Information: Diablo Canyon 50.69 Risk-Informed Categorization (EPID: L-2023-LLA-0137), dated July 2, 2024 (ML24184C042).

4.

U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018 (ML17317A256).

5.

U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 2, May 2011 (ML100910006).

6.

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, March 2009 (ML090410014).

7.

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, May 2006 (ML061090627).

8.

U.S. Nuclear Regulatory Commission, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, NUREG-1855, Revision 1, March 2017 (ML17062A466).

9.

U.S. Nuclear Regulatory Commission, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, NUREG-0800, Chapter 19, Section 19.2, June 2007 (ML071700658).

10.

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

11.

Nuclear Management and Resources Council, Inc., Guidelines for Industry Actions to Assess Shutdown Management, NUMARC 91-06, December 1991 (ML14365A203).

12.

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 (F&Os), dated February 21, 2017 (ML17086A431).

13.

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

14.

Nuclear Energy Institute, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, NEI 07-12, Revision 0, Draft H, November 2008 (ML083430464).

15.

American Nuclear Society, Fire PRA Methodology, ANSI/ANS 58.23-2007, November 2007 (withdrawn).

16.

U.S. Nuclear Regulatory Commission and Electric Power Research Institute, NUREG/CR-6850EPRI 1011989, Fire PRA Methodology for Nuclear Power Facilities, Volumes 1, Summary & Overview, and 2, Detailed Methodology September 30, 2005 (ML15167A401 and ML15167A411, respectively).

17.

American Society of Mechanical Engineers / American Nuclear Society, ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA S-2008, Standard for Level 1 /

Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, September 30, 2013.

18.

Peterson, D. B., Pacific Gas and Electric Company, letter to the U.S. Nuclear Regulatory Commission, License Amendment Request 23-01, Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505 [Technical Specifications Task Force (TSTF)], Revision 2, Provide Risk-Informed Extended Completion Times -

RITSTF [Risk-Informed TSTF] Initiative 4b, dated July 13, 2023 (ML23194A228).

19.

Peterson, D. B., Pacific Gas and Electric Company, letter to the U.S. Nuclear Regulatory Commission, Supplement to License Amendment Request 23-01, Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated June 15, 2024 (ML24016A299).

20.

Lee, S., U.S. Nuclear Regulatory Commission, letter to P. Gerfen, Pacific Gas and Electric Company, Diablo Canyon Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 245 and 247 Re: Revision to Technical Specifications to Adopt TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (EPID L-2023-LLA-0100)', dated May 29, 2024 (ML24099A219).

21.

Zoulis, A. M., U.S. Nuclear Regulatory Commission, memorandum to M. Franovich, U.S. Nuclear Regulatory Commission, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ML22014A084).

22.

Electric Power Research Institute, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI 1016737, December 2008.

23.

Markley, M. T., U.S. Nuclear Regulatory Commission, letter to Vice President, Operation, Entergy Operations, Inc., Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO-2 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 (TAC No. MD5250), dated April 22, 2009 (ML090930246).

24.

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Principal Contributors: Malcolm Patterson, NRR Hari Kodali, NRR Zeechung Wang, RES Fred Forsaty, NRR Mihaela Biro, NRR Steven Alferink, NRR Ming Li, NRR Gurjendra Bedi, NRR Stephen Cumblidge, NRR David Nold, NRR Dan Widrevitz, NRR Date: December 11, 2024

ML24269A083

• concurrence by email NRR-058 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DRA/APLA/BC* NRR/DRA/APLC/BC (A)*

NAME SLee PBlechman RPascarelli ANeuhausen DATE 9/24/2024 9/27/2024 9/11/2024 9/17/2024 OFFICE NRR/DEX/EEEB/BC*

NRR/DEX/EICB/BC*

NRR/DEX/EMIB/BC*

NRR/DSS/SCPB/BC*

NAME WMorton FSacko SBailey MValentin DATE 9/20/2024 9/18/2024 9/24/2024 9/16/2024 OFFICE NRR/DSS/SNSB/BC* NRR/DNRL/NPHP/BC* NRR/DNRL/NVIB/BC* OGC (NLO)*

NAME PSahd MMitchell ABuford (JTsao for)

MCarpenter DATE 9/12/2024 9/12/2024 9/12/2024 12/10/2024 OFFICE NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*

NAME TNakanishi SLee DATE 12/11/2024 12/11/2024