ML21137A247

Issuance of Amendment Nos. 233 and 230, Regarding Risk-Informed Categorization and Treatment of Structures, Systems, and Components

ML21137A247
Person / Time
Site:	Farley
Issue date:	06/30/2021
From:	Stephanie Devlin-Gill, Ed Miller Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Miller E, 415-2481
References
EPID L-2020-LLA-0134
Download: ML21137A247 (45)

Text

June 30, 2021 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENT NOS. 233 AND 230, TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2020-LLA-0134)

Dear Ms. Gayheart:

The Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 233 to Renewed Facility Operating License No. NPF-2 and Amendment No. 230 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant (FNP, Farley), Units 1 and 2, respectively. The amendments modify the licenses in response to your application dated June 18, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20170B114), as supplemented by letter dated March 2, 2021 (ADAMS Accession Nos. ML21064A526).

The amendments allow the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, and adds a license condition proposed by Southern Nuclear Operating Company, Inc. (SNC) to the renewed facility operating licenses that identifies action items that need to be completed prior to implementing 10 CFR 50.69 at Farley, Units 1 and 2 and identifies possible changes to the categorization process that would require prior NRC approval.

C.

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

Sincerely,

/RA/

G. Edward Miller, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 233 to NPF-2

2. Amendment No. 230 to NPF-8

3. Safety Evaluation cc: Listserv

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 233 Renewed License No. NPF-2

1.

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

A.

The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated June 18, 2020, as supplemented by letter dated March 2, 2021 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's regulations; D.

The issuance of this license 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 Commission's 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 paragraph C.(8) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:

(8) 10 CFR 50.69 Risk-Informed Categorization SNC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in SNC's submittal letter dated June 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 233 dated June 30, 2021.

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

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 30, 2021 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2021.06.30 11:18:13 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 233 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. NPF-2, page 4 License No. NPF-2, page 4 License No. NPF-2, page 9 License No. NPF-2, page 9 License No. NPF-2, page 10 License No. NPF-2, page 10 License No. NPF-2, page 11 License No. NPF-2, page 11 License No. NPF-2, page 12 License No. NPF-2, page 12 License No. NPF-2, page 13 License No. NPF-2, page 13

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 233 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 233, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3)

Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the Issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

a.

Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

b.

Deleted per Amendment 13

c.

Deleted per Amendment 2

d.

Deleted per Amendment 2

e.

Deleted per Amendment 152 Deleted per Amendment 2

f.

Deleted per Amendment 158

g.

Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1)

Identification of a sampling schedule for the critical parameters and control points for these parameters;

2)

Identification of the procedures used to quantify parameters that are critical to control points;

3)

Identification of process sampling points;

4)

A procedure for the recording and management of data;

5)

Procedures defining corrective actions for off control point chemistry conditions; and (b)

The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from February 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c)

The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously.

(8) 10 CFR 50.69 Risk-Informed Categorization SNC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in SNC's submittal letter dated June 18, 2020, and all its subsequent associated supplements as specified in License Amendment No.  dated -XQH.

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

D.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," and was submitted on May 15, 2006.

Farley - Unit 1 5HQHZHG/LFHQVH1R NPF-2 

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Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 186, as supplemented by a change approved by License Amendment No. 199.

E.

This renewed license is subject to the following additional conditions for the protection of the environment:

(1)

Southern Nuclear shall operate the facility within applicable Federal and State air and water quality standards and the Environmental Protection Plan (Appendix B).

(2)

Before engaging in an operational activity not evaluated by the Commission, Southern Nuclear will prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than evaluated in the Final Environmental Statement, Southern Nuclear shall provide a written evaluation of such activities and obtain prior approval of the Director, Office of Nuclear Reactor Regulation, for the activities.

F.

Alabama Power Company shall meet the following antitrust conditions:

(1)

Alabama Power Company shall recognize and accord to Alabama Electric Cooperative (AEC) the status of a competing electric utility in central and southern Alabama.

(2)

Alabama Power Company shall offer to sell to AEC an undivided ownership interest in Units 1 and 2 of the Farley Nuclear Plant. The percentage of ownership interest to be so offered shall be an amount based on the relative sizes of the respective peak loads of AEC and the Alabama Power Company (excluding from the Alabama Power Company's peak load that amount imposed by members of AEC upon the electric system of Alabama Power Company) occurring in 1976. The price to be paid by AEC for its proportionate share of Units 1 and 2, determined in accordance with the foregoing formula, will be established by the parties through good faith negotiations. The price shall be sufficient to fairly reimburse Alabama Power Company for the proportionate share of its total costs related to the Units 1 and 2 including, but not limited to, all costs of construction, installation, ownership and licensing, as of a date, to be agreed to by the two parties, which fairly accommodates both their respective interests. The offer by Alabama Power Company to sell an undivided ownership interest in Units 1 and 2 may be conditioned, at Alabama Power Company's option, on the agreement by AEC to waive any right of partition of the Farley Plant and to avoid interference in the day-to-day operation of the plant.

Farley - Unit 1 5HQHZHG/LFHQVH1R NPF-2 

$PHQGPHQW1R

(3)

Alabama Power Company will provide, under contractual arrangements between Alabama Power Company and AEC, transmission services via its electric system (a) from AEC's electric system to AEC's off-system members; and (b) to AEC's electric system from electric systems other than Alabama Power Company's and from AEC's electric system to electric systems other than Alabama Power Company's. The contractual arrangements covering such transmission services shall embrace rates and charges reflecting conventional accounting and ratemaking concepts followed by the Federal Energy Regulatory Commission (or its successor in function) in testing the reasonableness of rates and charges for transmission services. Such contractual arrangements shall contain provisions protecting Alabama Power Company against economic detriment resulting from transmission line or transmission losses associated therewith.

(4)

Alabama Power Company shall furnish such other bulk power supply services as are reasonably available from its system.

(5)

Alabama Power Company shall enter into appropriate contractual arrangements amending the 1972 Interconnection Agreement as last amended to provide for a reserve sharing arrangement between Alabama Power Company and AEC under which Alabama Power Company will provide reserve generating capacity in accordance with practices applicable to its responsibility to the operating companies of the Southern Company System. AEC shall maintain a minimum level expressed as a percentage of coincident peak one-hour kilowatt load equal to the percent reserve level similarly expressed for Alabama Power Company as determined by the Southern Company System under its minimum reserve criterion then in effect. Alabama Power Company shall provide to AEC such data as needed from time to time to demonstrate the basis for the need for such minimum reserve level.

(6)

Alabama Power Company shall refrain from taking any steps, including but not limited to, the adoption of restrictive provisions in rate filings or negotiated contracts for the sale of wholesale power, that serve to prevent any entity or group of entities engaged in the retail sale of firm electric power from fulfilling all or part of their bulk power requirements through self-generation or through purchases from some other source other than Alabama Power Company.

Alabama Power Company shall further, upon request and subject to reasonable terms and conditions, sell partial requirements power to any such entity. Nothing in this paragraph shall be construed as preventing an applicant from taking reasonable steps, in accord with general practice in the industry, to ensure that the reliability of its system is not endangered by any action called for herein.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No.

(7)

Alabama Power Company shall engage in wheeling for and at the request of any municipally-owned distribution system:

a.

of electric energy from delivery points of Alabama Power Company to said distribution system(s); and b.

of power generated by or available to a distribution system as a result of its ownership or entitlement2 in generating facilities, to delivery points of Alabama Power Company designated by the distribution system.

Such wheeling services shall be available with respect to any unused capacity on the transmission lines of Alabama Power Company, the use of which will not jeopardize Alabama Power Company's system. The contractual arrangements covering such wheeling services shall be determined in accordance with the principles set forth in Condition (3) herein.

Alabama Power Company shall make reasonable provisions for disclosed transmission requirements of any distribution system(s) in planning future transmission. "Disclosed" means the giving of reasonable advance notification of future requirements by said distribution system(s) utilizing wheeling services to be made available by Alabama Power Company.

(8)

The foregoing conditions shall be implemented in a manner consistent with the provisions of the Federal Power Act and the Alabama Public Utility laws and regulations thereunder and all rates, charges, services or practices in connection therewith are to be subject to the approval of regulatory agencies having jurisdiction over them.

Southern Nuclear shall not market or broker power or energy from Joseph M. Farley Nuclear Plant, Units 1 and 2. Alabama Power Company shall continue to be responsible for compliance with the obligations imposed on it by the antitrust conditions contained in this paragraph 2.F. of the renewed license. Alabama Power Company shall be responsible and accountable for the actions of its agent, Southern Nuclear, to the extent said agent's actions may, in any way, contravene the antitrust conditions of this paragraph 2.F.

G.

Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:

2 "Entitlement" includes, but is not limited to, power made available to an entity pursuant to an exchange agreement.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 

(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 (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing 2.

Dose to onsite responders H.

In accordance with the requirement imposed by the October 8, 1976 order of the United States Court of Appeals for the District of Columbia Circuit in Natural Resources Defense Council vs. Nuclear Regulatory Commission, No. 74-1385 and 74-1586, that the Nuclear Regulatory Commission "shall make any licenses granted between July 21, 1976 and such time when the mandate is issued subject to the outcome of such proceeding herein," this renewed license shall be subject to the outcome of such proceedings.

I.

This renewed operating license is effective as of the date of issuance and shall expire at midnight on June 25, 2037.

FOR THE NUCLEAR REGULATORY COMMISSION J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Appendix A - Technical Specifications 2.

Preoperational Tests, Startup Tests and Other Items Which Must Be Completed Prior to Proceeding to Succeeding Operational Modes 3.

Appendix B - Environmental Protection Plan 4.

Appendix C - Additional conditions Date of Issuance: May 12, 2005 Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 230 Renewed License No. NPF-8 1.

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

A.

The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated June 18, 2020, as supplemented by letter dated March 2, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's regulations; D.

The issuance of this license 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 Commission's 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 paragraph C.(25) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:

(25) 10 CFR 50.69 Risk-Informed Categorization:

SNC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in SNC's submittal letter dated June 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 230 dated June 30, 2021.

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

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 30, 2021 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2021.06.30 11:19:20 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 230 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. NPF-8, page 3 License No. NPF-8, page 3 License No. NPF-8, page 8 License No. NPF-8, page 8 License No. NPF-2, page 9 License No. NPF-2, page 9 License No. NPF-2, page 10 License No. NPF-2, page 10 License No. NPF-2, page 11 License No. NPF-2, page 11 License No. NPF-2, page 12 License No. NPF-2, page 12

Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 230 (2)

Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed license.

(3)

Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproducts, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporate below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 230, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3)

Delete per Amendment 144 (4)

Delete Per Amendment 149 (5)

Delete per Amend 144 Farley - Unit 2 Renewed License No.13)

Amendment No. 230 The Southern Nuclear Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. Southern Nuclear shall complete these activities no later than March 31, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.



Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must meet

the test procedures and reporting requirements of American Society for

Testing and Materials (ASTM) E 185-82 to the extent practicable for the

configuration of the specimens in the capsule. Any changes to the

capsule withdrawal schedule, including spare capsules, must be approved

by the NRC prior to implementation. All capsules placed in storage must

be maintained for future insertion.



10 CFR 50.69 Risk-Informed Categorization SNC is approved to implement 10 CFR 50.69 using the processes for

categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3,

and RISC-4 Structures, Systems, and Components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated

with internal events, including internal flooding, and internal fire; the

shutdown safety assessment process to assess shutdown risk; the

Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to

assess passive component risk for Class 2 and Class 3 and non-Class

SSCs and their associated supports; the results of the non-PRA

evaluations that are based on the IPEEE Screening Assessment for

External Hazards updated using the external hazard screening

significance process identified in ASME/ANS PRA Standard RA-Sa-2009

for other external hazards except seismic; and the alternative seismic

approach as described in SNC's submittal letter dated June 18, 2020, and

all its subsequent associated supplements as specified in License

Amendment No. 230 dated June 30, 2021.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the

categorization process specified above (e.g., change from a seismic

margins approach to a seismic probabilistic risk assessment approach).

D.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," and was submitted on May 15, 2006.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 181, as supplemented by a change approved by License Amendment No. 195.

E.

Deleted per Amendment 144 F.

Alabama Power Company shall meet the following antitrust conditions:

(1)

Alabama Power Company shall recognize and accord to Alabama Electric Cooperative (AEC) the status of a competing electric utility in central and southern Alabama.

(2)

Alabama Power Company shall offer to sell to AEC an undivided ownership interest in Units 1 and 2 of the Farley Nuclear Plant. The percentage of ownership interest to be so offered shall be an amount based on the relative sizes of the respective peak loads of AEC and Alabama Power Company (excluding from the Alabama Power Company's peak load that amount imposed by members of AEC upon the electric system of Alabama Power Company) occurring in 1976.

The price to be paid by AEC for its proportionate share of Units 1 and 2, determined in accordance with the foregoing formula, will be established by the parties through good faith negotiations. The price shall be sufficient to fairly reimburse Alabama Power Company for the proportionate share of its total costs related to the Units 1 and 2 including, but not limited to, all costs of construction, installation, ownership and licensing, as of a date, to be agreed to by the two parties, which fairly accommodates both their respective interests. The offer by Alabama Power Company to sell an undivided ownership interest in Units 1 and 2 may be conditioned, at Alabama Power Company's option, on the agreement by AEC to waive any right of partition of the Farley Plant and to avoid interference in the day-to-day operation of the plant.

(3)

Alabama Power Company will provide, under contractual arrangements between Alabama Power Company and AEC, transmission services via its electric system (a) from AEC's electric system to AEC's off-system members; and (b) to AEC's electric system from electric systems other than Alabama Power Company's, and from AEC's electric system to electric systems other than Alabama Power Company's. The contractual arrangements covering such transmission services shall embrace rates and charges reflecting conventional accounting and ratemaking concepts followed by the Federal Energy Regulatory Commission (or its successor in function) in testing the reasonableness of rates and charges for transmission services. Such contractual arrangements shall contain provisions protecting Alabama Power Company against economic detriment resulting from transmission line or transmission losses associated therewith.

Farley - Unit 2 Renewed License No.13)

Amendment No. 

(4)

Alabama Power Company shall furnish such other bulk power supply services as are reasonably available from its system.

(5)

Alabama Power Company shall enter into appropriate contractual arrangements amending the 1972 Interconnection Agreement as last amended to provide for a reserve sharing arrangement between Alabama Power Company and AEC under which Alabama Power Company will provide reserve generating capacity in accordance with practices applicable to its responsibility to the operating companies of the Southern Company System. AEC shall maintain a minimum level expressed as a percentage of coincident peak one-hour kilowatt load equal to the percent reserve level similarly expressed for Alabama Power Company as determined by the Southern Company System under its minimum reserve criterion then in effect. Alabama Power Company shall provide to AEC such data as needed from time to time to demonstrate the basis for the need for such minimum reserve level.

(6)

Alabama Power Company shall refrain from taking any steps, including but not limited, to the adoption of restrictive provisions in rate filings or negotiated contracts for the sale of wholesale power, that serve to prevent any entity or group of entities engaged in the retail sale of firm electric power from fulfilling all or part of their bulk power requirements through self-generation or through purchases from some other source other than Alabama Power Company.

Alabama Power Company shall further, upon request and subject to reasonable terms and conditions, sell partial requirements power to any such entity. Nothing in this paragraph shall be construed as preventing an applicant from taking reasonable steps, in accord with general practice in the industry, to ensure that the reliability of its system is not endangered by any action called for herein.

(7)

Alabama Power Company shall engage in wheeling for and at the request of any municipally-owned distribution system:

a.

of electric energy from delivery points of Alabama Power Company to said distribution system(s); and b.

of power generated by or available to a distribution system as a result of its ownership or entitlement2 in generating facilities, to delivery points of Alabama Power Company designated by the distribution system.

2 "Entitlement" includes, but is not limited to, power made available to an entity pursuant to an exchange agreement.

Farley - Unit 2 Renewed License No.13)

Amendment No. 

Such wheeling services shall be available with respect to any unused capacity on the transmission lines of Alabama Power Company, the use of which will not jeopardize Alabama Power Companys system. The contractual arrangements covering such wheeling services shall be determined in accordance with the principles set forth in Condition (3) herein.

Alabama Power Company shall make reasonable provisions for disclosed transmission requirements of any distribution system(s) in planning future transmission. "Disclosed" means the giving of reasonable advance notification of future requirements by said distribution system(s) utilizing wheeling services to be made available by Alabama Power Company.

(8)

The foregoing conditions shall be implemented in a manner consistent with the provisions of the Federal Power Act and the Alabama Public Utility laws and regulations thereunder and all rates, charges, services or practices in connection therewith are to be subject to the approval of regulatory agencies having jurisdiction over them.

Southern Nuclear shall not market or broker power or energy from Joseph M. Farley Nuclear Plant, Units 1 and 2. Alabama Power Company shall continue to be responsible for compliance with the obligations imposed on it by the antitrust conditions contained in this paragraph 2.F. of the renewed license. Alabama Power Company shall be responsible and accountable for the actions of its agent, Southern Nuclear, to the extent said agent's actions may, in any way, contravene the antitrust conditions of this paragraph 2.F.

G.

The facility requires relief from certain requirements of 10 CFR 50.55a(g) and exemptions from Appendices G, H and J to 10 CFR Part 50. The relief and exemptions are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 5. They are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, the relief and exemptions are hereby granted.

With the granting of these relief and exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

H.

Southern Nuclear shall immediately notify the NRC of any accident at this facility which could result in an unplanned release of quantities of fission products in excess of allowable limits for normal operation established by the Commission.

Farley - Unit 2 Renewed License No.13)

Amendment No. 

I.

Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions 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 (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing 2.

Dose to onsite responders J.

Alabama Power Company 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.

K.

This renewed operating license is effective as of the date of issuance and shall expire at midnight on March 31, 2041.

FOR THE NUCLEAR REGULATORY COMMISSION J. E. Dyer, Director Office of Nuclear Reactor Regulation

Attachment:

1.

Appendix A - Technical Specifications (NUREG-0697, as revised) 2.

Appendix B - Environmental Protection Plan 3.

Appendix C - Additional conditions Date of Issuance: May 12, 2005 Farley - Unit 2 Renewed License No.13)

Amendment No. 

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 233 TO FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 230 TO FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated June 18, 2020 (Reference 1), as supplemented by letter dated March 2, 2021 (Reference 2), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a license amendment request (LAR) for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP, Farley). The proposed amendments would modify the Farley licensing basis (LB) by adding a license condition to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

In e-mail correspondence dated January 12, 2021 (Reference 3), the NRC staff requested additional information (RAI) from the licensee. The licensee responded to the RAIs in its supplement dated March 2, 2021. The licensees supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 11, 2020 (85 FR 48571).

The licensee proposed the following license condition to the Farley renewed facility operating licenses to allow the implementation of 10 CFR 50.69:

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

[Probabilistic Risk Assessment] Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in SNC's submittal letter dated June 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. [233 and 230] dated June 30, 2021.

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

The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated and systematic risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance. 1

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design basis functions.

For SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.

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

SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.

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

1 RG 1.201, Revision 2, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, describes the SSC categorization process as an approach that includes both PRA and non-probabilistic methods. (Reference 6) 2.2 Regulatory Guidance The NRC staff considered the following regulatory guidance during its review of the proposed changes:

x 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) x RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 5) x RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference 6) x NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis:

General Guidance (Reference 7) x NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference 8) x Federal Register notice (69 FR 68008, 68028-68029; November 22, 2004) related to Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Industry Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 9), as endorsed by RG 1.201, for trial use with clarifications and describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

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

x Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

x Sections 3, 4, 5, and 7 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

x Section 6 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

x Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).

x Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

x Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).

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

2.3 Previous NRC Approvals Reviewed by NRC Staff Arkansas Nuclear One, Unit 2, Alternative Request A plant-specific alternative request for Arkansas Nuclear One, Unit 2 (ANO-2), included a risk-informed repair/replacement application (RI-RRA) to determine risk-informed safety classification and treatment program for repair and replacement activities for Class 2 and Class 3 pressure-retaining items or their associated supports (exclusive of Class CC and MC items). Entergy proposed a safety classification methodology that differs from ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1, (Reference 10), but the NRC The NRC reviewed the proposed request, in part, based on consistency of the proposed changes to N-660 with the requirements of 10 CFR 50.69, with categorization guidelines in NEI 00-04 as endorsed in RG 1.201, and approved the ANO-2 plant-specific alternative by letter dated April 22, 2009, which included the NRC staffs safety evaluation (SE) (Reference 11) that described NRC authorization for classification of Class 2 and 3 passive components..

Calvert Cliffs Nuclear Power Plant, Units 1 and 2, License Amendment Request NRC approved a plant-specific alternative seismic approach in its LAR for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (Calvert Cliffs) dated February 28, 2020 (Reference 12), as a Tier 1 site with low seismic hazard and high seismic margin. This precedent SE was based, in part, on case studies in Electric Power Research Institute (EPRI) Report 3002012988 (Reference 13).

The NRC staff considered the following criteria for the applicability and use of the proposed seismic Tier 1 approach:

x Ground motion response spectrum (GMRS) peak acceleration is at or below approximately 0.2 ground acceleration (g) or where the GMRS is below or approximately equal to the safe shutdown earthquake (SSE) between 1.0 hertz (Hz) and 10 Hz.

The licensee provided a supplemental letter dated March 2, 2021., In that letter, SNC stated that the technical criteria in EPRI Report 3002017583 (Reference 14), Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, submitted for Farley is unchanged from EPRI Report 3002012988 and that EPRI Report 3002017583 is updated from EPRI Report 3002012988 by the incorporation of RAI responses into EPRI Report 3002012988 from the review of the Calvert Cliffs 50.69 LAR. The NRC staffs review of Farleys use of the alternative seismic approach in EPRI Report 3002017583 is provided in Section 3.3.1.2 of this SE.

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed LB changes, including both permanent and temporary changes, is to show that the proposed LB changes meet the five key principles stated in Section C of RG 1.174. These key principles are:

Principle 1:

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

Principle 2:

The proposed LB change is consistent with the defense-in-depth (DID) philosophy.

Principle 3:

The proposed LB change maintains sufficient safety margins.

Principle 4:

When the proposed LB change results in an increase in risk, the increase 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 LB change should be monitored using performance measurement strategies.

3.2 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles of RG 1.174, Revision 3 and is pertinent to: (1) compliance with current regulations, (2) evaluation of DID, and (3) evaluation of safety margins.

3.2.1 Key Principle 1: Licensing Basis Change Meets the Current Regulations 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 into one of the following four RISC categories, which are defined in 10 CFR 50.69(a), as follows:

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 are classified as having either HSS functions (i.e., RISC-1 and RISC-2 categories) or LSS functions (i.e., RISC-3 and RISC-4 categories). For HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements 2 NEI 00-04, 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 (i.e., SSCs that are categorized as RISC-1 or RISC-2), as used in 10 CFR 50.69.

from these SSCs). For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For RISC-3 SSCs, licensees can replace special treatment with an alternative treatment. For RISC-4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.

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 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 The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1, 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 SE, used the framework provided in RG 1.174, Revision 3, and NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1.

Section 2 of NEI 00-04, Revision 0, in part, states that the categorization process includes eight primary steps:

1. Assembly of Plant-Specific Inputs (Section 3 of NEI 00-04, Revision 0)

2. System Engineering Assessment (Section 4 of NEI 00-04, Revision 0)

3. Component Safety Significance Assessment (Section 5 of NEI 00-04, Revision 0)

4. Defense-In-Depth Assessment (Section 6 of NEI 00-04, Revision 0)

5. Preliminary Engineering Categorization of Functions (Section 7 of NEI 00-04, Revision 0)

6. Risk Sensitivity Study (Section 8 of NEI 00-04, Revision 0)

7. Integrated Decisionmaking Panel Review and Approval (Section 9 of NEI 00-04, Revision 0)

8. SSC Categorization (Section 10 of NEI 00-04, Revision 0)

In Section 3.1 of the Enclosure to its letter dated June 18, 2020, the licensee stated that it will implement the risk-informed categorization process in accordance with NEI 00-04, as endorsed in RG 1.201. In Section 3.2.3 of the Enclosure to its letter dated June 18, 2020, the licensee proposed to use a risk-informed graded approach that meets the requirements of 10 CFR 50.69 (b)(2) as an alternative to those listed in NEI 00-04 section 1.5 and 5.3 for its seismic hazard assessment. The licensee stated this approach is specified in EPRI Report 3002017583. The NRC notes that use of this alternative method is a deviation from the endorsed industry guidance. A more detailed NRC staff review of alternative methods is provided in Section 3.3.1.2 of this SE.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, and the monitoring outlined in NEI 00-04, Revision 0, and clarifications in RG 1.201, Revision 1, ensures that the SSC categorization process is sufficient to assure that the SSC functions continue to be met and that any performance deficiencies will be identified and appropriate corrective actions taken. The licensees SSC categorization program includes the appropriate steps/elements prescribed in NEI 00-04, Revision 0, to assure that SSCs specified are appropriately categorized consistent with 10 CFR 50.69. Based on the above, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision making prescribed in RG 1.174, Revision 3.

3.2.2 Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth Philosophy In RG 1.174, Revision 3, the NRC identified the following considerations used for evaluating the impact of the LB change on DID:

x Preserve a reasonable balance among the layers of defense.

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

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

x Preserve adequate defense against potential common-cause failures.

x Maintain multiple fission product barriers.

x Preserve sufficient defense against human errors.

x Continue to meet the intent of the plants design criteria.

RG 1.201, Revision 1, endorses the guidance in Section 6 of NEI 00-04 (Reference [9]), but notes that the containment isolation criteria in this section of the guidance are separate and distinct from those set forth in 10 CFR 50.69(b)(1)(x). The criteria in 10 CFR 50.69(b)(1)(x) are to be used in determining which containment penetrations and valves may be exempted from the Type B and Type C leakage testing requirements in both Options A and B of Appendix J to 10 CFR Part 50. The criteria provided in paragraph 50.69(b)(1)(x) of 10 CFR are not to determine the proper RISC category for containment isolation valves or penetrations.

In Section 3.1.1 of the Enclosure to its letter dated June 18, 2020, the licensee clarified that it will require an SSC to be categorized as HSS based on the DID assessment performed in accordance with NEI 00-04. Based on the above, the NRC staff concludes that the proposed change is consistent with the DID philosophy described in key principle 2 of RG 1.174, and is, therefore, acceptable. The NRC staff finds that the licensee's process is consistent with the NRC-endorsed NEI guidance and would meet the 10 CFR 50.69(c)(1)(iii) criterion that requires DID to be maintained.

3.2.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The regulations in 10CFR50.69(c)(1)(iv) requires the evaluations to provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from changes in treatment are small. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of the SSC to assure that sufficient safety margins are maintained. With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the LB (e.g.,

FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data. RG 1.174, Revision 3 (Reference [4]) provides guidelines for making that assessment including evaluations to ensure the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The SSCs design basis functions as described in the plants LB, including the Updated Final Safety Analysis Report and Technical Specifications Bases, do not change and should continue to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant LB. Based on the above, the NRC staff concludes that the licensee has established a program to ensure sufficient safety margins are maintained in accordance with the third key principle of RG 1.174, Revision 3 and would therefore meet 10 CFR 50.69(c)(1)(iv).

3.3 Risk-Informed Assessment 3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04, Revision 0, endorsed by RG 1.201, Revision 1, addresses the fourth and fifth key principles of RG 1.174, Revision 3, pertaining to the assessment for change in risk.

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

In its letter dated June 18, 2020, the licensee described that the Farley categorization process uses PRA-modeled hazards to assess risks for the internal events (includes internal flood and internal fires). The SSC categorization process uses the following non-PRA methods to assess risks from the other risk contributors:

x Seismic hazard: Alternative seismic treatment using guidance from EPRI Report 3002017583 and qualitative insights about seismic risk at Farley.

x External hazards: Screening analysis performed for IPEEE (Reference 15) updated using criteria from Part 6 of the ASME/ANS RA-Sa-2009 - Addenda A to RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, the PRA Standard, (Reference 16).

x Other Hazards: Screening analysis performed for the IPEEE updated using criteria from Part 6 of the PRA Standard.

x Shutdown Events: Safe Shutdown Risk Management program consistent with NUMARC 91-06 (Reference 17).

x Passive Components: ANO-2 passive categorization methodology.

With the exception of seismic, the approaches and methods proposed by the licensee to address internal events, external events, other hazards, DID, and shutdown events are consistent with the approaches and methods included in the guidance in NEI 00-04, Revision 0.

To address seismic hazard in the SSC categorization process, the licensee proposed to use an alternative method not endorsed by the NRC. A detailed NRC staff review of the licensees proposed alternative seismic approach is provided in Section 3.3.1.2 of this SE. The non-PRA method for the categorization for passive components is consistent with the ANO-2 methodology for passive components approved for risk-informed safety classification and treatment for repair/replacement activities in Class 2 and 3 moderate-and high-energy systems.

The use of the ANO-2 methodology in the SSC categorization process is discussed in Section 3.3.1.2 of this SE.

3.3.1.1 Scope of the PRA The Farley PRA is comprised of a full-power, Level 1, internal events (IEPRA) and a fire PRA (FPRA) to evaluate the Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) risk metrics. The licensee discussed in Section 3.3 of the Enclosure to its letter dated June 18, 2020, that the internal events PRA model (which includes internal floods) has been assessed against RG 1.200. Furthermore, in Section 3.3 of the Enclosure to its letter dated June 18, 2020, the licensee states that a finding closure review was conducted on the identified PRA model in October 2018. Open findings were reviewed and closed using the NRC-accepted process (Reference 18) documented in the NEI letter to the NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-13, 'Close-out of Facts and Observations, dated February 21, 2017 (Reference 19).

The NRC staff evaluated the scope of the PRA including: (1) peer-review history and results, (2) the independent assessment of fact and observation (F&O) closure, (3) credit for FLEX in the PRA, and (4) assessment of assumptions and approximations. The staff's review of these aspects of the PRA and supplemental responses to assess for consistency with the applicable processes as endorsed by the NRC, where necessary, are provided below.

Internal Events PRA (Including Internal Floods) Peer-Review History In Section 3.3 of the Enclosure to its letter dated June 18, 2020, the licensee stated that the IEPRA model was subjected to a self-assessment and full-scope peer review in March 2010. It was assessed against the supporting requirements of the PRA Standard at Capability Category II (CC II). Subsequently, in October 2018, SNC conducted an independent assessment for closure of the finding-level F&Os and concluded all the IEPRA (includes internal floods) F&Os have been closed. An NRC staff review of this independent assessment is provided below in this section of the SE.

In Section 3.2 of the Enclosure to its letter dated June 18, 2020, for the PRAs, SNC stated, in part, that there are no PRA upgrades that have not been peer reviewed. In Section 3.3 of the Enclosure to its letter dated June 18, 2020, and in response to NRC staffs RAI dated March 2, 2021, the licensee summarized significant IEPRA (includes internal floods), FPRA, and SPRA model changes that triggered a focused-scope peer review, which was conducted in December of 2019. The review resulted in no finding level F&Os.

Therefore, the NRC staff concludes that the Farley internal events PRA (including internal floods) was peer reviewed appropriately, consistent with RG 1.200, and the F&Os have been adequately resolved to assess the impact on the risk-informed application.

Internal Fire PRA Peer Review History In Section 3.2.2 of the Enclosure to its letter dated June 18, 2020, the licensee stated that the Farley categorization process for fire hazards will use a peer reviewed plant-specific FPRA model and that the fire PRA model was developed consistent with NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, and only utilizes methods previously accepted by the NRC. The licensee's fire PRA was subject to a full-scope industry peer review in October 2011, consistent with RG 1.200 against supporting requirements of the ASME/ANS PRA Standard at CC II. In February and July 2018, focused-scope peer reviews were performed to address supporting requirements that were not met at CC II during the 2011 peer review.

The finding-level F&Os from the 2011 full-scope and the two 2018 focused-scope reviews were considered fully resolved by the independent assessment review team in September 2018.

Therefore, in accordance with RG 1.200, no F&Os associated with the fire PRA were provided in the Enclosure to its letter dated June 18, 2020.

Based on the above, the NRC staff concludes that the Farley fire PRA was peer reviewed appropriately, consistent with RG 1.200, and the F&Os have been adequately dispositioned to assess the impact on the risk-informed application.

Appendix X, Independent Assessment Process for F&O Closure Section X.1.3 of Appendix X to NEI 05-04/07-12/12-13 (Reference 19) provides guidance to perform an independent assessment for the closure of F&O identified from a full-scope or focused-scope peer review.

In Section 3.2 of the Enclosure to its letter dated June 18, 2020, for all the PRA models, SNC states in part, there are no PRA upgrades that have not been peer reviewed.Section X.1.3 of Appendix X to NEI 05-04/07-12/12-13, as accepted by NRC staff in a memorandum dated May 3, 2017, provides guidance that includes a written assessment and justification of whether the resolution of each F&O, within the scope of the independent assessment, constitutes a PRA upgrade or maintenance update as defined in the PRA Standard. Therefore, the NRC staff concludes that there are no additional peer reviews required.

The NRC staff concludes that the Independent Assessments performed in December 2018 and November 2019 were performed consistent with Appendix X to NEI 05-04/07-12/12-13.

Credit for FLEX Equipment The NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (Reference 20), provides the NRC staffs assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decision making in accordance with the guidance of RG 1.200, Revision 2.

In the RAI dated January 12, 2021, the NRC noted that Section 3.3 of the Enclosure to a letter dated June 18, 2020, indicated that FLEX equipment and actions have been credited in the PRA models and given the significant challenges of modeling FLEX equipment and actions without sufficient industry data and without a consensus approach for human reliability analysis (HRA) to address unique aspects of FLEX actions. NRC staff requested identification of the FLEX equipment and actions that were credited in the internal events and fire PRA models and description about how they were modeled. In response to the RAI, the licensee performed a sensitivity by raising FLEX equipment failure probabilities by a factor of five, where a factor of three would result in sensitivity larger than the base case 95th percentile. The sensitivity analysis showed that there was a small change to CDF and LERF and, therefore, no significant impact on the 50.69 program risk assessment. Based on the above, the NRC staff finds the licensees modelling of FLEX in the PRAs for this application acceptable because the licensees sensitivity analysis showed negligible impact for this application.

Identification and Treatment of Key Assumptions and Sources of Uncertainty Farley confirmed that it used the guidance in NUREG-1855 to identify, screen, and characterize the sources of model uncertainty and related assumptions in the base PRA that are relevant to this application. Sub-step E-1.4 of the guidance is a qualitative screening process that involves identifying and validating whether consensus models have been used in the PRA to evaluate identified model uncertainties.3 The licensee confirmed that some uncertainties and assumptions were screened based on the use of a consensus method for the Farley uncertainty analysis. The licensee presented identified key assumptions and sources of uncertainty in the Enclosure to its letter dated June 18, 2020. The NRC staff finds that the assessment performed to identify the key assumptions/sources of uncertainty is consistent with the guidance provided in NUREG-1855, Revision 1.

3 Per 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.

The licensee provided dispositions to the key assumptions and sources of uncertainty it identified in the Enclosure to its letter dated June 18, 2020. The licensee provided analysis to determine each source had minimal impact on the application. The licensee concluded that no other sensitivity studies beyond those mandated by NEI 00-04 are required for the 10 CFR 50.69 program.

In Section 3.2.7 of the Enclosure to its letter dated June 18, 2020, the licensee confirmed that sensitivity studies will be performed consistent with Table 5-2 of the industry guidance. In accordance with Section 9 of NEI 00-04 as endorsed by RG 1.201, the licensees integrated decision-making panel (IDP) will use information and risk insights compiled in the initial categorization process, including awareness of the limitations and assumptions of the PRA, and combines that with other information from design bases, DID, and safety margins to finalize the categorization of the SSCs. The NRC staff finds it sufficient that the licensees plan to perform a sensitivity study consistent with Table 5-2 of NEI 00-04 and the provision for the IDP to consider PRA assumptions and simplifications during the SSC categorization process to address the identified key assumptions and sources of uncertainty in the context of the decision making under consideration for the categorization of the SSC at the time of the risk analysis being performed.

The NRC staff recognizes that 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).

Paragraph 50.69(e) and (f) stipulates the process for feedback and adjustment to assure configuration control is maintained for these routine changes and updates to the PRA(s).

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 and fire PRA to support SSC categorization is endorsed by RG 1.201 (Reference [6]). The PRAs must be acceptable to support the categorization process and must be subjected to a peer-review process assessed against a standard that is endorsed by the NRC. Revision 2 of RG 1.200 provides guidance for determining the acceptability of the PRA by comparing the PRA to the relevant parts of the PRA Standard (Reference [13]) using a peer-review process.

The licensee has subjected the internal events PRA and fire PRA to the peer-review processes and submitted the results of the peer review. The NRC staff reviewed the peer-review history (which included the results and findings), the licensee's resolution of peer-review findings, and the identification and disposition of key assumptions and sources of uncertainty. The NRC staff concludes that (1) the licensee's IEPRA and FPRA are acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201, Revision 1, and (2) the key assumptions for the PRAs have been identified consistent with the guidance in RG 1.200, Revision 2, and NUREG-1855, as applicable, and appropriately addressed for this application.

In the RAI dated January 12, 2021, the NRC staff noted that RG 1.174 provided risk acceptance guidelines in terms of total core damage or large early release frequency (CDF and LERF respectively). The Farley total CDF of 8.4X10-05/reactor-year approaches this acceptance guideline. In response to the RAI dated March 2, 2021, the licensee provided mean calculated CDF and LERF risk values which demonstrated that the Farley application meets the criteria of the RG, albeit still close to the acceptance guidelines.

The NRC staff finds the licensee provided sufficient information, and the internal events PRA (includes internal floods) and fire PRA are acceptable, and therefore, meet the requirements set forth in 50.69(c)(1)(i) and (ii) of 10 CFR 50.69.

3.3.1.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization Alternative Seismic Approach As part of its proposed process to categorize SSCs according to safety significance, the licensee proposed to use a non-PRA method to consider seismic hazards. The regulations in 10 CFR 50.69(c)(1)(ii) and 50.69(b)(2)(ii) permit the use of systematic evaluation techniques in the risk-informed categorization process. The licensee provided a description of its proposed alternative seismic approach for considering seismic risk in the categorization process and described how it would be used in the categorization process in Section 3.2.3 of the Enclosure to its letter dated June 18, 2020, and its supplement dated March 2, 2021. The licensee based its plant-specific evaluation, in part, on the case studies performed in EPRI Report 3002017583.

The licensee stated that the case studies are applicable to Farley and are used in the alternative seismic approach. In its supplement dated March 2, 2021, the licensee cited a previous approval relevant to its proposed alternative seismic approach. The licensee identified plant-specific differences between Farley and the NRC staffs previous approval for Calvert Cliffs Nuclear Plant SE dated February 28, 2020. The information presented in the licensees submittal and supplement, as well as in EPRI Report 3002017583, provided sufficient detail for the NRC staff to evaluate the licensees proposed alternative seismic approach, how the licensees proposed alternative seismic approach would be used in the categorization process; and the measures for assuring the quality and level of detail for the licensees proposed alternative seismic approach for the categorization of SSCs. Based on the above, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(ii) for the proposed alternative seismic approach would be met.

EPRI Report 3002017583 includes the results from case studies performed to determine the extent and type of unique high-safety-significant (HSS) SSCs from seismic PRAs (SPRAs). In its supplement dated March 2, 2021, the licensee indicated that aside from updates included in an RAI submittal for the Calvert Cliffs 50.69 LAR into the previous version of this report, EPRI 3002012988, the technical criteria in EPRI Report 3002017583 are unchanged from its predecessor report EPRI Report 3002012988. The NRC staffs review confirmed that the case studies in EPRI Report 3002017583 used by the licensee to support its proposed alternative seismic approach, as well as the information in its supplements, provided sufficient plant-specific evaluation of the applicability and differences for Farley as compared to the previous approval of the Calvert Cliffs SE dated February 28, 2020. The information presented in the letter dated June 18, 2020, and its supplement dated March 2, 2021, provided a sufficient description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv) for the alternative seismic approach. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed alternative seismic approach.

The licensee also explained that there are two differences between the licensees proposed alternative seismic approach and the alternative seismic approach previously approved for Calvert Cliffs. The first is an updated Figure 1-2 to replace the EPRI Report 3002017583 guidance document and replacement of Figure 2-2, Low Seismic Hazard Site: Typical SSE to GMRS Comparison, since the original versions figure was incorrect. With regard to the Calvert Cliffs RAI 4.b response in the July 1, 2019, supplement (Reference 21), Farley stated the plant

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has a RG 1.200, Revision 2, peer-reviewed seismic PRA hazard model with all F&Os closed.

However, the licensee will use the alternative seismic Tier 1 approach, instead of their SPRA, for the categorization process in this LAR. The NRC staff finds this acceptable because the technical criteria used in EPRI Report 3002017583, as docketed in the licensees supplement dated March 2, 2021, is sufficient to support the NRC staffs acceptance of the Farley plant-specific alternative seismic approach.

Evaluation of Technical Acceptability of the PRAs Used for Case Studies Supporting the Proposed Alternative Seismic Approach In its supplement dated March 2, 2021, the licensee responded to the NRC staffs requests for information concerning the previous approval including the case studies, mapping approach, and conclusions on the determination of unique HSS SSCs from the case studies which were used by the licensee to support its proposed alternative seismic approach. The licensee stated that the Calvert Cliffs supplemental responses to RAIs associated with the case study Plants A, C, and D, pertaining to the technical acceptability of the PRAs used, as well as the technical adequacy of certain technical details of the conduct of the case studies, were incorporated into EPRI 3002017583 and are applicable to Farley. The licensees submittal does not address plant-specific applicability to Farley for Plant B trial categorization for reactor coolant pump seals as a PRA upgrade described in EPRI report 3002017583. The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the case studies for Plants A, C, and D in EPRI Report 3002017583 and the licensees assertion of plant-specific applicability of the previous approval. The NRC staff also evaluated the peer review process and resolution of peer-review findings and key assumptions and sources of uncertainties for Plants A, C, and D.

Based on the above, the NRC staff finds that the acceptability of PRAs used in the Plant A, C, and D case studies in EPRI report 3002017583, the mapping approach used in those case studies, and the conclusions on the determination of unique HSS SSCs from the case studies in the previous approval are applicable to this licensees proposed plant-specific alternative seismic approach. Therefore, the NRC staff concludes that thelicensees plant-specific evaluation of Plant A, C, and D trial categorization is sufficient to support applicability of the proposed alternative seismic approach; the mapping of SSCs between the SPRA, the full-power IEPRA and, as applicable, the FPRA for the Plant A, C, and D case studies. The NRC concludes that the licensees plant-specific evaluation is technically justifiable to support conclusions on the determination of unique HSS SSCs from SPRAs in Plant A, C, and D case studies in EPRI Report 3002017583, and applicable to Farley and the licensees proposed alternative seismic approach.

Evaluation of the Criteria for the Proposed Alternative Seismic Approach In the Enclosure to its letter dated June 18, 2020, the licensee states that the GMRS peak acceleration for Farley is at or below approximately 0.2g, or where the GMRS is below or approximately equal to the SSE between 1.0 Hz and 10 Hz.

The licensee further stated that the GMRS-to-SSE comparison demonstrates that Farley qualifies as a Tier 1 plant under the criteria in EPRI report 3002017583 and that this comparison confirms the expected seismic risk at Farley would be very low. The NRC staff notes that the licensees plant-specific evaluation is supported by its NRC 10 CFR 50.54(f) response dated March 31, 2014 (Reference 22).

The NRC staff reviewed the licensees submittal and supplements and plant-specific evaluation and concludes that the proposed criteria in EPRI Report 3002017583 to determine the applicability and use of the proposed alternative seismic Tier 1 approach is acceptable.

Evaluation of Applicability of Criteria for this Application In Section 3.2.3 of the Enclosure to its June 18, 2020, letter, the licensee compared the Farley GMRS from the reevaluated seismic hazard developed and submitted by the licensee in response to Near-Term Task Force Recommendation 2.1 against the sites design-basis SSE, as shown in Figure A4-1 of the Enclosure to its letter dated June 18, 2020, to demonstrate that the site meets the criteria for application of the proposed alternative seismic approach as a Tier 1 plant. In Section 3.2.3 of the Enclosure to its June 18, 2020 letter, the licensee stated that the NRC staff concluded that the methodology used by the licensee in determining the GMRS was acceptable and that the GMRS determined by the licensee adequately characterized the reevaluated hazard for the Farley site. The NRC staffs review confirmed the licensees statements and the comparison of the GMRS from the reevaluated seismic hazard against the SSE. Based on its review, the NRC staff finds that the licensees seismic hazard meets the criteria for the proposed alternative seismic approach.

In Section 3.2.3 of the Enclosure to its letter dated June 18, 2020, the licensee stated that the small percentage contribution of seismic to total plant risk makes it unlikely that an integral importance assessment for a component, as defined in NEI 00-04, would result in an overall HSS determination.

The NRC staff reviewed the seismic risk estimates provided in Farleys risk-informed technical specifications Initiative 4b LAR submittal (Reference 23) as well as other plant risks determined from internal events and internal fire to support its review of this LAR. The NRC staff noted that seismic CDF is low compared to CDFs from internal events and internal fire. Further as noted in the NRC staffs SE for Farleys 4b LAR (Reference 24), the licensees seismic large early release frequency estimate in the 4b LAR is expected to be conservative because it was estimated using a fragility value appreciably lower than the fragility of components leading to containment isolation failure, such as the estimates in Appendix B of NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants (Reference 25).

Contemporary seismic PRAs for various licensees at different sites that have been reviewed by the NRC staff further support this determination. Based on the above, the NRC staff concludes that the seismic risk contributions for Farley, Units 1 and 2, are not expected to solely result in an SSC being categorized as HSS.

The NRC staff finds that the licensees basis for applying the proposed alternative seismic approach to its site is acceptable because: (1) the reevaluated hazard meets the criteria for use of the proposed alternative seismic approach, and (2) the seismic risk contribution would not solely result in a SSC being categorized as HSS.

Evaluation of the Implementation of Conclusions from the Case Studies The categorization conclusions from EPRI Report 3002017583 case studies, performed for GMRS to SSE ratios significantly higher than Farley, indicated that seismic-specific failure modes resulted in HSS categorization uniquely from SPRAs. Therefore, such seismic-specific failure modes, such as correlated failures, relay chatter, and passive component structural failure mode, can influence the categorization process. The NRC staff reviewed the proposed alternative seismic approach to evaluate whether the categorization-related conclusions from EPRI Report 3002017583 were included and implemented appropriately.

In Section 3.2.3 of the Enclosure to its letter dated June 18, 2020, the licensee discussed its proposed plant-specific alternative seismic approach. The licensee stated that the proposed categorization approach for seismic hazards will include qualitative consideration of the mitigation capabilities of SSCs during seismically-induced events and seismic failure modes, based on insights obtained from prior seismic evaluations performed for Farley.

The licensee explained that the qualitative characterization of seismic risk performed for the independent decision-making panel will include information from the various post-Fukushima seismic reviews including results of seismic walkdowns, seismic mitigation strategy assessment, and seismic high frequency evaluations. The objective of the alternative seismic approach, as described in Figure 3-1 in the Enclosure to LAR of June 18, 2020, is to identify plant-specific seismic insights derived from the components in the system being categorized.

In its supplement dated March 2, 2021, the licensee stated that its plant-specific evaluation considered differences in the proposed alternative seismic approach between Farley and the alternative seismic approach previously reviewed and approved by the NRC staff in the Calvert Cliffs SE dated February 28, 2020. The NRC staffs review of the licensees proposed alternative seismic approach determined that the previous approval is applicable to this licensees proposed alternative seismic approach and that the plant-specific evaluation of the implementation of the alternative seismic approach is acceptable. The NRC staffs review of the proposed alternative seismic approach, in conjunction with the requirements in 10 CFR 50.69 and the corresponding statement of consideration, finds that the proposed alternative seismic approach includes the evaluations required by 10 CFR 50.69(c)(1)(ii) as well as 10 CFR 50.69(c)(1)(iv) because:

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

2. The proposed alternative seismic approach presents system-specific seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, thereby providing the IDP a means to consider potential impacts of seismic events in the categorization process.

3. The insights presented to the IDP include potentially important seismically-induced failure modes, as well as mitigation capabilities of SSCs during seismically-induced design basis and severe accident events consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI Report 3002017583. The insights will use prior plant-specific seismic evaluations, and therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration. Further, the recommendation for categorizing civil structures in the alternative seismic approach provides appropriate consideration of such failures from a seismic event.

4. The proposed alternative seismic approach presents the IDP with the basis for the proposed alternative seismic approach, including the low seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach.

5. The proposed alternative seismic approach includes qualitative consideration and insights related to the impact of a seismic event on SSCs for each SSC that is categorized and does not limit the scope to SSCs from the case studies supporting this application.

Consideration of Changes to Seismic Hazard An important input to the NRC staffs evaluation of the proposed alternative seismic approach is the current knowledge of the seismic hazard at the plant. The possibility exists for the seismic hazard at the site to increase such that the criteria for use of the proposed alternative seismic approach are challenged. In such a situation, the categorization process may be impacted from a seismic risk perspective either solely due to the seismic risk or by the integrated importance measure determination.

In Section 3.2.3 of the Enclosure to its letter dated June 18, 2020, the licensee stated that U.S.

nuclear power plants that utilize the 10 CFR 50.69 Seismic Alternative (EPRI 3002017583) will continue to compare GMRS to SSE. Since the alternative seismic approach explicitly cites and is based on EPRI Report 3002017583, the continued comparison of GMRS to SSE applies to Farley. The licensee also stated that the seismic hazard at the plant is subject to periodic reconsideration as new information became available through industry evaluations.

The NRC staffs review finds that consideration of changes to the seismic hazard in the licensees plant-specific proposed alternative seismic approach considered the precedent appropriately. The NRC staffs evaluation of the consideration of changes to the seismic hazard against the requirements in 10 CFR 50.69(e)(1), 10 CFR 50.69(e)(3), and 10 CFR 50.69(d)(2)(ii) as well as the resulting conclusion on consideration of changes to the seismic hazard in the previous approval is applicable to this licensees proposed alternative seismic approach.

Consequently, the NRC staff finds that the consideration of changes to the seismic hazard at Farley that exceed the criteria for use of the proposed alternative seismic approach is acceptable for the proposed approach because: (1) the criteria for use of the proposed alternative seismic approach are clear and traceable, (2) the proposed alternative seismic approach includes periodic reconsideration of the seismic hazard as new information becomes available, (3) the proposed alternative seismic approach satisfies the requirements in 10 CFR 50.69 discussed above, and (4) the licensee has included a proposed license condition in the LAR to require NRC approval for a change to the specified seismic categorization approach.

Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach In Section 3.5 of the Enclosure to its letter dated June 18, 2020 and the March 2, 2021 supplement, the licensee stated that its configuration control process ensures that changes to the plant, including a physical change and changes to documents, are evaluated to determine the impact to drawings, design bases, licensing documents, programs, procedures, and training.

Further, the licensee cited an applicable NRC-approved precedent for its proposed alternative seismic approach.

The NRC staff found that consideration of the feedback and adjustment process in the licensees proposed alternative seismic approach is acceptable. The NRC staff finds that (1) the licensees programs provide reasonable assurance that the existing seismic capacity of Low Safety Significant components would not be significantly impacted, and (2) the monitoring and configuration control program ensures that potential degradation of the seismic capacity would be detected and addressed before significantly impacting the plant risk profile. Therefore, the NRC staff finds that the potential impact of the seismic hazard on the categorization is maintained acceptably low and the requirements in 10 CFR 50.69(c)(1)(iv) are met for the proposed alternative seismic approach.

Method for Assessing Other Non-Seismic External Hazards This hazard category includes all non-seismic external hazards such as high winds, external floods, transportation, and nearby facility accidents, and other hazards. The licensee discussed its consideration of other external hazards in Section 3.2.4 of the Enclosure to its letter dated June 18, 2020. The licensee stated that all other external hazards, except for seismic, were screened for applicability to Farley per a plant-specific evaluation in accordance with Generic Letter 88-20 and updated to use the criteria in ASME/ANS PRA Standard RA-Sa-2009.

In the Enclosure to its letter dated June 18, 2020, the licensee provided the results of the plant-specific evaluation that assessed the IPEEE results using endorsed criteria in the ASME/ANS RA-Sa-2009 PRA Standard and current plant hazard information. The NRC notes, that neither this plant-specific evaluation nor its results were peer reviewed against Part 6 of the ASME/ANS Ra-SA-2009 PRA Standard as endorsed in RG 1.200, Revision 2.

In irs supplement dated March 2, 2021, SNC stated its categorization assessment process will be consistent with the flow chart provided in Figure 5-6 in Section 5.4 of NEI 00-04.

The NRC staff finds that the bases for screening the other external hazards are acceptable because the licensee considers the as-built, as-operated plant, uses current hazard information, and will be updated if changes that can impact categorization are identified from periodic reviews. In summary, the use of the Farley IPEEE results described by the licensee in its letter dated June 18, 2020, its supplement dated March 2, 2021, and the licensee's assessment of the other external hazards (i.e., high winds, tornadoes, and external flood) is consistent with Section 5 of NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1. The NRC staff concludes that the licensee's treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

Method for Assessing Passive Components In Section 3.1.2 of the Enclosure to its letter dated June 18, 2020, the licensee proposed using a categorization method for passive components that was not approved for use in NEI 00-04 or RG 1.201, Revision 2, but was a plant-specific alternative approved for use by the NRC for ANO-2 on April 22, 2009, in the NRC staffs SE (Reference 11). The ANO-2 RI-RRA precedent is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1.

The ANO-2 plant-specific approval relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance is generally measured by the frequency and the consequence of, in this case, pipe ruptures. Treatment requirements (including repair/replacement) only affect the frequency of passive component failure.

Categorizing solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization is not generally affected by changes in frequency arising from changes to the treatment. Therefore, the NRC staff found for ANO that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.

In Section 3.1.2 of the Enclosure to its letter dated June 18, 2020, the licensee stated, The passive categorization process is intended to apply the same risk-informed process accepted in the ANO 2-R&R-004 for the passive categorization of Class 2, 3, and non-class components.

The licensee also stated that consistent with ANO 2-R&R-004, All ASME Code Class 1 SSCs with a pressure retaining function, as well supports, will be assigned high safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. The NRC staff finds that the licensee's proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

3.3.1.3 Key Principle 4 Conclusions Based on review of the scope and documented quality of the internal events PRA (which includes internal floods) and the fire PRA, taken together with non-PRA methods, the NRC staff concludes that the change in risk related to adoption of the proposed amendment is consistent with the Commissions safety goals. Therefore, it satisfies the fourth key principle for risk informed decision making prescribed in RG 1.174, Revision 3.

3.3.2 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04 provides guidance that includes programmatic configuration control and a periodic review to ensure that all aspects of the 10 CFR 50.69 program (i.e., 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 over time are continually incorporated.

Sections 11 and 12 of NEI 00-04, Revision 0, includes discussion on periodic review and program documentation and change control. Maintaining change control and periodic review will also maintain confidence that all aspects of the 10 CFR 50.69 program and risk categorization for SSCs continually reflect the Farley as-built, as-operated plant.

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

4.0 TECHNICAL CONCLUSION In the June 18, 2020, LAR, as supplemented by letter dated March 2, 2021, the licensee proposed to add a license condition to the Farley renewed facility operating licenses to allow the implementation of 10 CFR 50.69. Based on the NRC staffs review of the licensees LAR and response to RAIs, the NRC staff identified specific actions, as described below, that are identified as being necessary to support the NRC staffs conclusion that the proposed program meets the requirements in 10 CFR 50.69, the guidance in RG 1.201, Revision 1, and NEI 00-04, Revision 0. Additional actions (e.g., final procedures) have not been submitted by the licensee or reviewed by the NRC staff for issuance of this SE, but will be completed before implementation of the program as specified in the 10 CFR 50.69 rule.

The NRC staffs finding on the acceptability of the SSC categorization process and PRA evaluation in the licensees proposed 10 CFR 50.69 program is conditioned upon the license condition provided below. For the clarifications to the NEI 00-04, Revision 0 guidance and other changes that were described by the licensee, the NRC staff finds them to be routine and systematically addressed through the configuration management and control and periodic update processes as described in Sections 3.3.1.1 and 3.3.2 of this SE.

The licensee proposed to add Paragraph C.(8) to Renewed Facility Operating License No. NPF-2 and Paragraph C.(25) to Renewed Facility Operating License No. NPF-8 for Farley, Unit Nos. 1 and 2, respectively. The proposed license condition states:

SNC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in SNC's submittal letter dated June 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. [233] dated June 30, 2021.

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

The NRC staff finds that the proposed license condition is acceptable, because: (1) it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that previously has been endorsed by the NRC; and (2) the evaluation in SE Section 3.3.1.2, finds the non-PRA methods for assessing risk for internal fires, seismic, and passive components, which are deviations from NEI 00-04, to be acceptable.

The NRC staff notes that the guidance for implementing 10 CFR 50.69 provided by the Commission in the Federal Register notice dated November 22, 2004 (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 TSs (including Improved Technical Specifications (ITS) and the associated Technical Requirements Manual (TRM)) to be part of the 10 CFR 50.69 rule. Therefore, the licensee should continue to follow its TSs (including the ITS and TRM, as applicable) when implementing a 10 CFR 50.69 license amendment.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the NRC notified the State of Alabama official of the proposed issuance of the amendment on April 28, 2021. On May 17, 2021, the State official confirmed that the State of Alabama had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in Title 10 of the Code of Federal Regulations (10 CFR) Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there have been no public comments on such finding published in the Federal Register (85 FR 48571; August 11, 2020). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

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

8.0 REFERENCES

1 Gayheart, Cheryl A., Southern Nuclear Operating Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Joseph M. Farley Nuclear Plant, Units 1 and 2 -

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," June 18, 2020 (ADAMS Accession No. ML20170B114).

2 Gayheart, Cheryl A., Southern Nuclear Operating Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Joseph M. Farley Nuclear Plant, Units 1 and 2 - Response to Request for Additional Information Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," March 2, 2021 (ADAMS Accession No. ML21064A526).

3 U. S. Nuclear Regulatory Commission (NRC), Email to Wesley A. Sparkman and Ryan M.

Joyce, Southern Nuclear Operating Co., Inc., "Request for Additional Information - Joseph M. Farley Nuclear Plant, Units 1 and 2 - Application to Adopt 10 CFR 50.69 "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," January 12, 2021 (ADAMS Accession No. ML21012A324).

4 U. S. Nuclear Regulatory Commission (NRC), "Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision on Plant-Specific Changes to the Licensing Basis," January 2018 (ADAMS Accession No. ML17317A256).

5 U. S. Nuclear Regulatory Commission (NRC), "Regulatory Guide (RG) 1.200, Revision 2 -

An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009 (ADAMS Accession No. ML090410014).

6 U. S. Nuclear Regulatory Commission (NRC), "Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," May 2006 (ADAMS Accession No. ML061090627).

7 U. S. Nuclear Regulatory Commission (NRC), "NUREG-0800, SRP for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 19, Section 19-2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis General Guidance," June 2007 (ADAMS Accession No. ML071700658).

8 U. S. Nuclear Regulatory Commission (NRC); Brookhaven National Laboratory and Sandia National Laboratories, "NUREG-1855, Revision 1 - Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," March 2017 (ADAMS Accession No. ML17062A466).

9 Nuclear Energy Institute (NEI), "NEI 00-04, Rev. 0 - 10 CFR 50.69 SSC [Structures, Systems and Components] Categorization Guideline," July 2005 (ADAMS Accession No. ML052910035).

10 American Society of Mechanical Engineers (ASME), "Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI.

Division 1," July 2002.

11 U. S. Nuclear Regulatory Commission (NRC), letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems," April 22, 2009 (ADAMS Accession No. ML090930246).

12 U. S. Nuclear Regulatory Commission (NRC), letter to Bryan C. Hanson, Exelon Generation Company, LLC, "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," February 28, 2020 (ADAMS Accession No. ML19330D909).

13 Electric Power Research Institute (EPRI), "Report No. 3002012988 - Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization,"

Palo Alto, CA, July 2018.

14 Electric Power Research Institute (EPRI), "Report No. 3002017583 - Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization,"

February 2020 (Enclosure 2 of ADAMS Accession No. ML21064A526).

15 U. S. Nuclear Regulatory Commission (NRC), "Generic Letter (GL) 88-20, Supplement 4 -

Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities

- 10 CFR 50-54(f)," June 28, 1991 (ADAMS Accession No. ML031150485).

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

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

17 Nuclear Management and Resources Council, Inc. (NUMARC), "NUMARC 91 Guidelines for Industry Actions to Assess Shutdown Management," December 1991 (ADAMS Accession No. ML14365A203).

18 U. S. Nuclear Regulatory Commission (NRC), letter to Nuclear Energy Institute (NEI), "U. S.

Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Closeout of Facts and Observations (F&Os)," May 3, 2017 (ADAMS Accession No. ML17079A427).

19 Nuclear Energy Institute (NEI), "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Closeout of Facts and Observations (F&Os)," February 21, 2017 (ADAMS Accession No. ML17086A431).

20 U. S. Nuclear Regulatory Commission (NRC), Internal Memo, "Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis," May 30, 2017 (ADAMS Accession No. ML17031A269).

21 Exelon Generation, LLC., letter to U. S. Nuclear Regulatory Commission (NRC), "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.59, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," July 1, 2019 (ADAMS Accession No. ML19183A012).

22 Southern Nuclear Operating Company, Inc., letter to U. S. Nuclear Regulatory Commission (NRC), "Joseph M. Farley Nuclear Plant, Units 1 and 2 - Seismic Hazard and Screening Report for CEUS Sites," March 31, 2014 (ADAMS Accession No. ML14092A020).

23 Southern Nuclear Operating Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Joseph M. Farley Nuclear Plant, Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS),"

Guidelines." July 27, 2018 (ADAMS Accession No. ML18208A619).

24 U. S. Nuclear Regulatory Commission (NRC), letter to Southern Nuclear Operating Company, "Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos.

225 and 222 Regarding Implementation of NEI 06-09, Risk-Informed Technical Specifications Initative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A," August 23, 2019 (ADAMS Accession No. ML19175A243).

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

"NUREG/CR-4334 - An Approach to the Quantification of Seismic Margins in Nuclear Power Plants," August 1985 (ADAMS Accession No. ML090500182).

Principal Contributors:

Malcolm Patterson, NRR De Wu, NRR Date issued: June 30, 2021

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