Text

Issuance of Amendments Regarding Revision to Technical Specifications to Adopt TSTF-554, Revise RCS Leakage
	ML21349A518
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	02/02/2022
From:	John Lamb; NRC/NRR/DORL/LPL2-1
To:	Gayheart C; Southern Nuclear Operating Co
	Lamb J
References
	EPID L-2021-LLA-0175
	Download: ML21349A518 (49)
	v • d • e

February 2, 2022 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; EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2; AND VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2021-LLA-0175)

Dear Ms. Gayheart:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 240 to Renewed Facility Operating License No. NPF-2 and Amendment No. 237 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively; Amendment No. 314 to Renewed Facility Operating License No. DPR-57 and Amendment No. 259 to Renewed Facility Operating License No. NPF-5 for the Edwin I. Hatch Nuclear Plant, Unit Nos 1 and 2, respectively; and Amendment No. 213 to Renewed Facility Operating License NPF-68 and Amendment No. 196 to Renewed Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the License and Technical Specifications (TSs) in response to your application dated September 29, 2021.

The proposed amendments would revise the TSs to adopt TSTF-554, Revise Reactor Coolant Leakage Requirements.

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.

If you have questions, you can contact me at 301-415-3100 or John.Lamb@nrc.gov.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348, 50-364, 50-321, 50-366, 50-424, and 50-425

Enclosures:

1. Amendment No. 240 to NPF-2

2. Amendment No. 237 to NPF-8 3 Amendment No. 314 to DPR-57

4. Amendment No. 259 to NPF-5

5. Amendment No. 213 to NPF-68

6. Amendment No. 196 to NPF-81

7. 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. 240 Renewed License No. NPF-2

1.

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

A.

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

(Southern Nuclear), dated September 29, 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 to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 240, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 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: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:48:00 -05'00'

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. 237 Renewed License No. NPF-8

1.

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

A.

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

(Southern Nuclear), dated September 29, 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 to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 237, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 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: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:49:05 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 240 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND ATTACHMENT TO LICENSE AMENDMENT NO. 237 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 License and Appendix A Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain vertical 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-8, page 3 License No. NPF-8, page 3 TSs TSs 1.1-3 1.1-3 1.1-4 1.1-4 3.4.13-1 3.4.13-1 3.4.13-2 3.4.13-2

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 240 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 240, 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

Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 237 (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 2821 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 237, 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

Definitions 1.1 (continued)

Farley Units 1 and 2 1.1-3 Amendment No. 240 (Unit 1)

Amendment No. 237 (Unit 2) 1.1 Definitions AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

Definitions 1.1 Farley Units 1 and 2 1.1-4 Amendment No. 240 (Unit 1)

Amendment No. 237 (Unit 2) 1.1 Definitions LEAKAGE

3.

Reactor Coolant System (RCS) LEAKAGE (continued) through a steam generator (SG) to the Secondary System;

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14, Initial Tests and Operation, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

RCS Operational LEAKAGE 3.4.13 Farley Units 1 and 2 3.4.13-1 Amendment No. 240 (Unit 1)

Amendment No. 237 (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

RCS Operational LEAKAGE 3.4.13 Farley Units 1 and 2 3.4.13-2 Amendment No. 240 (Unit 1)

Amendment No. 237 (Unit 2)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2

NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

NOTES--------------------------------

1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

NOTE--------

Only required to be performed during steady state operation In accordance with the Surveillance Frequency Control Program SR 3.4.13.2

NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 150 gallons per day through any one SG.

In accordance with the Surveillance Frequency Control Program

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 314 Renewed License No. DPR-57

1.

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

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 29, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 314, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date 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 Renewed Facility Operating License No. DPR-57 and Technical Specifications Date of Issuance: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:51:29 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 314 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) 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. DPR-57, page 4 License No. DPR-57, page 4 TSs TSs 1.1-4 1.1-4 3.4-7 3.4-7 3.4-8 3.4-8 Renewed License No. DPR-57 Amendment No. 314 for sample analysis or instrument calibration, or associated with radioactive apparatus or components (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 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 following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal.

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 314, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

(3)

Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Definitions 1.1 HATCH UNIT 1 1.1-4 Amendment No. 314 1.1 Definitions (continued)

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval RECIRCULATION from initial signal generation by the associated turbine stop valve limit PUMP TRIP switch or from when the turbine control valve hydraulic control (EOC-RPT) oil pressure drops below the pressure switch setpoint to complete SYSTEM suppression of the electric arc between the fully open contacts of the RESPONSE TIME recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE The INSERVICE TESTING PROGRAM is the licensee program that fulfills TESTING the requirements of 10 CFR 50.55a(f).

PROGRAM LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems;

b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation GENERATION in an arbitrary length of fuel rod, usually six inches. It is the integral of the RATE heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL required logic components (i.e., all required relays and contacts, trip TEST units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

(continued)

RCS Operational LEAKAGE 3.4.4 HATCH UNIT 1 3.4-7 Amendment No. 314 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

5 gpm unidentified LEAKAGE;

c.

30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; and

d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C.

Unidentified LEAKAGE increase not within limit.

C.1 Reduce LEAKAGE increase to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.4 HATCH UNIT 1 3.4-8 Amendment No. 314 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. NPF-5

1.

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

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 29, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-5 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 259 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date 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 Renewed Facility Operating License No. NPF-5 and Technical Specifications Date of Issuance: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:53:03 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 259 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) 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-5, page 4 License No. NPF-5, page 4 TSs TSs 1.1-4 1.1-4 1.1-5 1.1-5 3.4-7 3.4-7 3.4-8 3.4-8

Renewed License No. NPF-5 Amendment No. 259 (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 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 following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 259 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(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 license supported by a favorable evaluation by the Commission.

(a) Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would 2 The original licensee authorized to possess, use, and operate the facility with Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Definitions 1.1 HATCH UNIT 2 1.1-4 Amendment No. 259 1.1 Definitions (continued)

EMERGENCY The ECCS RESPONSE TIME shall be that time interval from when CORE COOLING the monitored parameter exceeds its ECCS initiation setpoint at the SYSTEM (ECCS) channel sensor until the ECCS equipment is capable of performing

RESPONSE

its safety function (i.e., the valves travel to their required positions, TIME pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval RECIRCULATION from initial signal generation by the associated turbine stop valve limit PUMP TRIP switch or from when the turbine control valve hydraulic control oil (EOC-RPT) pressure drops below the pressure switch setpoint to complete SYSTEM suppression of the electric arc between the fully open contacts of the RESPONSE TIME recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE The INSERVICE TESTING PROGRAM is the licensee program that TESTING fulfills the requirements of 10 CFR 50.55a(f).

PROGRAM ISOLATION The ISOLATION SYSTEM RESPONSE TIME shall be that time interval SYSTEM from when the monitored parameter exceeds its isolation initiation RESPONSE TIME setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; (continued)

Definitions 1.1 HATCH UNIT 2 1.1-5 Amendment No. 259 1.1 Definitions LEAKAGE c.

Total LEAKAGE (continued)

Sum of the identified and unidentified LEAKAGE; and d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation GENERATION in an arbitrary length of fuel rod, usually six inches. It is the integral of the RATE heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL required logic components (i.e., all required relays and contacts, trip TEST units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM The MCPR shall be the smallest critical power ratio (CPR) that CRITICAL POWER exists in the core for each class of fuel. The CPR is that power RATIO (MCPR) in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE -

A system, subsystem, division, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

(continued)

RCS Operational LEAKAGE 3.4.4 HATCH UNIT 2 3.4-7 Amendment No. 259 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

5 gpm unidentified LEAKAGE; c.

30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; and d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C.

Unidentified LEAKAGE increase not within limit.

C.1 Reduce LEAKAGE increase to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.4 HATCH UNIT 2 3.4-8 Amendment No. 259 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 213 Renewed License No. NPF-68

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 29, 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 as 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 set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-68 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 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 License No. NPF-68 and the Technical Specifications Date of Issuance: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:54:37 -05'00'

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 196 Renewed License No. NPF-81

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 29, 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 as 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 set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-81 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 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 License No. NPF-81 and the Technical Specifications Date of Issuance: February 2, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.02.02 14:55:46 -05'00'

ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 TO LICENSE AMENDMENT NO. 213 RENEWED FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 196 RENEWED FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) 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-68, page 4 License No. NPF-68, page 4 License No. NPF-81, page 3 License No. NPF-81, page 3 TSs TSs 1.1-3 1.1-3 1.1-4 1.1-4 3.4.13-1 3.4.13-1 3.4.13-2 3.4.13-2 3.4.13-3 Renewed Operating License NPF-68 Amendment No. 213 (1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4)

Deleted (5)

Deleted (6)

Deleted (7)

Deleted (8)

Deleted (9)

Deleted (10)

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

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training and 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 Renewed Operating License NPF-81 Amendment No. 196 (2)

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this 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 byproduct, 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; (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 my be produced by the operation of the facility authorized herein.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 1 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 incorporated below.

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be

Definitions 1.1 (continued)

Vogtle Units 1 and 2 1.1-3 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2) 1.1 Definitions (continued)

- AVERAGE shall be the average (weighted in proportion to DISINTEGRATION ENERGY the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 14 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE (ESF) RESPONSE interval from when the monitored parameter exceeds its TIME ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

Definitions 1.1 (continued)

Vogtle Units 1 and 2 1.1-4 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2) 1.1 Definitions LEAKAGE

b.

Unidentified LEAKAGE (continued)

All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

RCS Operational LEAKAGE 3.4.13 Vogtle Units 1 and 2 3.4.13-1 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.13 Vogtle Units 1 and 2 3.4.13-2 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2

NOTE-------------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

NOTES--------------------------

1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Only required to be performed during steady state operation.

3. Not applicable to primary to secondary LEAKAGE.

Perform RCS water inventory balance.

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving steady state operation AND In accordance with the Surveillance Frequency Control Program (continued)

RCS Operational LEAKAGE 3.4.13 Vogtle Units 1 and 2 3.4.13-3 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.13.2

NOTE---------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 150 gallons per day through any one SG.

In accordance with the Surveillance Frequency Control Program

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NPF-2 AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NPF-8 AMENDMENT NO. 314 TO RENEWED FACILITY OPERATING LICENSE DPR-57 AMENDMENT NO. 259 TO RENEWED FACILITY OPERATING LICENSE NPF-5 AMENDMENT NO. 213 TO RENEWED FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 196 TO RENEWED FACILITY OPERATING LICENSE NPF-81 JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NOS. 50-348, 50-364, 50-321, 50-366, 50-424 AND 50-425

1.0 INTRODUCTION

By application dated September 29, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21273A072), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for amendments to the Technical Specifications (TSs) for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2; Edwin I.

Hatch Nuclear Plant (Hatch), Units 1 and 2; and Vogtle Electric Generating Plant (Vogtle),

Units 1 and 2 (collectively, the SNC Operating Fleet).

In its application dated September 29, 2021, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed LAR under the Consolidated Line Item Improvement Process (CLIIP).

The proposed changes would revise the TSs related to reactor coolant system (RCS) operational leakage and the definition of the term LEAKAGE based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, (TSTF-554) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20016A233), and the associated NRC staff safety evaluation (SE) of TSTF-554 (ADAMS Accession No. ML20322A024).

The licensee has proposed variations from the TS changes described in Traveler TSTF-554, Revision 1. The variations are described and evaluated in Section 3.2 of this SE.

1.1 Reactor Coolant System Description Components that contain or transport the coolant to or from the reactor core make up the RCS.

Materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and through operational wear or mechanical deterioration during normal and upset operating environments. Such material degradation could lead to leakage of reactor coolant into containment buildings.

The RCS leakage falls under two main categories - identified leakage and unidentified leakage.

Identifying the sources of leakage is necessary for prompt identification of potentially adverse conditions, assessment of safety significance of the leakage, and quick corrective action. A limited amount of leakage from the reactor coolant pressure boundary (RCPB) directly into the containment/drywell atmosphere is expected as the RCS and other connected systems cannot be made 100 percent leak tight. This leakage is detected, located, and isolated from the containment atmosphere so as to not interfere with measurement of unexpected RCS leakage detection.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Separation of identified leakage from unidentified leakage provides quantitative information to the operators, allowing them to take corrective action should leakage occur that is detrimental to the safety of the unit and the public.

1.2 Proposed TS Changes to Adopt TSTF-554 In accordance with NRC staff-approved TSTF-554, the licensee proposed changes that would revise the TSs related to RCS operational leakage and the definition of the term LEAKAGE.

Specifically, the licensee proposed the following TS changes to adopt TSTF-554:

For each facility, the TS 1.1 identified LEAKAGE definition a.2 would be revised to remove the exclusion of pressure boundary leakage from identified leakage by deleting either and the phrase not to be pressure boundary LEAKAGE.

Pressure boundary leakage is currently defined in the TSs as leakage (except primary to secondary leakage in a pressurized-water reactor) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. For each facility, the TS 1.1 pressure boundary LEAKAGE definition would be revised to delete the word nonisolable. The sentence, LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE, would be relocated from the Technical Specifications Bases and added to the definition. This change would revise paragraph c. of the LEAKAGE definition in

TS 1.1 for Farley, Units 1 and 2; Vogtle; Units 1 and 2 and paragraph d. of the LEAKAGE definition in TS 1.1 for Hatch, Units 1 and 2.

Additionally, for each facility, the LEAKAGE definition would be revised by other editorial and punctuation changes to reflect the deletion and listed definitions.

The RCS operational LEAKAGE requirements in TS 3.4.4 for Hatch Units and TS 3.4.13 for Farley and Vogtle Units, would be revised as follows:

o For each facility, a new Condition A to isolate the pressure boundary leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would be added to the ACTIONS section.

o Existing Conditions A, B, and C in TSs for Hatch Units and existing Conditions A and B in TSs for Farley and Vogtle Units, and the associated required actions would be renumbered.

o Existing Condition C in TSs for Hatch Units and existing Condition B in TSs for Farley and Vogtle Units, would be revised to delete the condition Pressure boundary LEAKAGE exists, because pressure boundary leakage would be addressed by the new Condition A. In addition, this Condition would be revised to be applicable when the required action and associated completion time of the other conditions are not met.

1.3 Additional Proposed TS Changes The application identified certain variations from TSTF-554. Section 3.2 of this SE provides a description of the variations and the Staffs evaluation.

2.0 REGULATORY EVALUATION

The regulation at 10 CFR 50.36(c)(2) requires that TSs include limiting conditions for operation (LCOs). Per 10 CFR 50.36(c)(2)(i), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The regulation at 10 CFR 50.2 defines the RCPB as all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves, which are: (1) [p]art of the reactor coolant system, or (2) [c]onnected to the reactor coolant system Regulatory Guide (RG) 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, dated May 2008 (ADAMS Accession No. ML073200271), Section B, Discussion Leakage Separation, provides information related to separation between identified and unidentified leakage.

The NRC staff has provided guidance for the review of TSs in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP),

March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed

changes are consistent with NUREG 14311 and NUREG 14332, as modified by NRC-approved travelers. Traveler TSTF-554 revised the STSs related to RCS operational leakage and the definition of the term LEAKAGE. The NRC approved TSTF-554, as a CLIIP traveler on December 18, 2020 (ADAMS Package Accession No. ML20324A083).

TSTF-554, Revision 1, and its corresponding Safety Evaluation discuss the applicable regulatory requirements and guidance, including the applicability of 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 14 and 30.

Vogtle, Units 1 and 2, Farley, Units 1 and 2, and Hatch, Unit 2, were licensed to the 10 CFR Part 50, Appendix A, GDC. However, Hatch, Unit 1, was not licensed to the 10 CFR Part 50, Appendix A, GDC. Hatch, Unit 1, was licensed to the applicable Atomic Energy Commission (AEC) preliminary general design criteria identified in Federal Register a t 32 FR 10213, published July 11, 1967 (ADAMS Accession No. ML043310029). The applicable AEC proposed criteria were generally comparable to the 10 CFR Part 50, Appendix A, GDC, as documented in the Hatch Updated Final Safety Analysis Report (UFSAR), Appendix F, Conformance to the Atomic Energy Commission (AEC) Criteria.

TSTF-554 references 10 CFR Part 50, Appendix A, GDC 14, Reactor coolant pressure boundary, which states:

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

TSTF-554 references 10 CFR Part 50, Appendix A, GDC 30, Quality of reactor coolant pressure boundary, which states:

Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical.

Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Design evaluation related to GDC 14 and 30 are included in Section F.3 of the Hatch, Unit 1, UFSAR, Evaluation with Respect to 1971 General Design Criteria. The Enclosure to the licensees letter dated September 29, 2021, stated, Following implementation of the proposed change Hatch, Unit 1, will remain in compliance with applicable AEC design criteria as described in the Hatch, Unit 1, UFSAR. Therefore, this difference does not alter the conclusion that the proposed TSTF-554 is applicable to Hatch, Unit 1.

1 U. S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 4, dated April 2012 (ADAMS Accession No. ML12100A222 and ML12100A228, respectively).

2 U. S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR Plants, NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0 dated April 2012 (ADAMS Accession No. ML12104A192 and ML12104A193, respectively).

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-554 The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-554. In accordance with the SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-554 are applicable to Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2, TSs because the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are pressurized-water reactors (PWRs), and Hatch, Units 1 and 2, are boiling-water reactors (BWRs), and the NRC staff approved the TSTF-554 changes for PWR and BWR designs. The NRC finds that the licensees proposed changes to the Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2, TSs in Section 1.1 of this SE are consistent with those found acceptable in TSTF-554.

In the SE of TSTF-554, the NRC staff concluded that TSTF-554 changes to the STS 1.1 definition of LEAKAGE and to the STS for RCS operational leakage (the LCO addressing conditions and required actions when RCS pressure boundary leakage exists), are acceptable.

The NRC staff found that removing the term nonisolable provides a clearer definition of pressure boundary leakage and that the source of the leakage is not relevant to this capability provided that separate, appropriate limits on pressure boundary leakage have been established.

Therefore, the proposed change to the definition of identified leakage was acceptable as it did not conflict with 10 CFR 50.2 and was consistent with RG 1.45. The NRC staff further found that proposed new Condition A on boundary pressure leakage, including its associated Required Action A.1 and Completion Time, is acceptable because the LCO revisions continue to specify the lowest functionable capability of equipment, identify remedial actions, and require shutdown of the reactor if the remedial actions cannot be met.

The NRC staff finds that proposed changes to the TS definition for each facility clarify what constitutes pressure boundary leakage, and the source of leakage does not matter if the TSs have separate limits on pressure boundary leakage. The NRC staff also finds that TS 3.4.4 for Hatch Units and TS 3.4.13 for Farley and Vogtle Units correctly specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. Also, the NRC staff finds that proposed changes to the Actions of LCO 3.4.13 are adequate remedial actions to be taken until each LCO can be met to provide protection to the health and safety of the public. Thus, the proposed changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i) as discussed in Section 3.0 of the NRC staffs SE for TSTF-554.

3.2 Technical Evaluation of Additional Proposed TS Changes 3.2.1 Editorial Changes The licensee noted that Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2, TSs have different numbering and nomenclature than standard technical specifications (STSs),

as follows.

Farley, Units 1 and 2, and Vogtle, Units 1 and 2: Item 3 under Definition Identified LEAKAGE already ends with a ; so the markups in this LAR editorially differ from TSTF-554, which shows the punctuation changing from a. to a ;.

Farley, Units 1 and 2: The definition Pressure Boundary LEAKAGE in the Farley, Units 1 and 2, TS uses the term SG LEAKAGE rather than primary to secondary

LEAKAGE. Revision 3 of TSTF-449 included an editorial change to replace except SG LEAKAGE with except primary to secondary LEAKAGE. The editorial change resolved an inconsistency between the definition and the TS (e.g., Farley, Units 1 and 2, TS 3.4.13 CONDITION A). This difference is due to Farley, Units 1 and 2, implementing a prior revision of TSTF-449, which did not yet include this editorial change.

For consistency with the STS and TS 3.4.13, this L A R is requesting this editorial change to the definition.

Hatch, Units 1 and 2: Item a.1 under definition Identified LEAKAGE has a comma in it to separate the appositive such as that from pump seals or valve packing from the rest of the sentence. Neither TSTF-554 nor the STS have this comma. This LAR does not request a change to this.

Hatch, Units 1 and 2: Items a.1 and a.2 under definition Identified LEAKAGE already end with semicolons so the markups in this LAR editorially differ from TSTF-554, which shows the punctuation changing from a, to a ;.

Hatch, Units 1 and 2, and Vogtle, Units 1 and 2: Item b Unidentified LEAKAGE under Definition LEAKAGE already ends with a ; so the markups in this L A R editorially differ from TSTF-554 which shows the punctuation changing from a. to a ;.

Hatch, Units 1 and 2: Item c. Total LEAKAGE under Definition LEAKAGE does not have an and at the end; Hatch, Units 1 and 2, are adding this and with this LAR for editorial consistency with the TSTF-554 markup and the STS.

The NRC staffs review finds the above variations acceptable since the differences are editorial and do not affect the applicability of TSTF-554 to the proposed LAR.

3.2.2 Other Variations In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations.

The TSs for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, TS 3.4.13, Required Action B.2 (revised to C.2), direct the plants to be in Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and are modified by a Note that states that LCO 3.0.4.a is not applicable when entering Mode 4.

The TSTF-554 markup, Required Action B.2 (revised to C.2), requires the plants to be in Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This difference is due to incorporation of TSTF-432, Revision 1, Change in Technical Specifications End States (WCAP-16294), (ADAMS Accession Nos. ML103360003 and ML100770146) into the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, TSs. TSTF-432 was not incorporated into Revision 4 of the STS on which TSTF-554 is based. The licensees letter dated September 29, 2021, stated, This difference does not affect the applicability of TSTF-554 to the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, TSs.

The STS markup of TS 3.4.4, RCS Operational LEAKAGE, in TSTF-554 renames Required Actions B.1 and B.2 to Required Actions C.1 and C.2. STS Required Action B.2 states, Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel. Hatch, Units 1 and 2, Required Action B.1 is renamed C.1, but the Hatch, Units 1 and 2, TSs do not contain an equivalent to

Required Action B.2. The licensees letter dated September 29, 2021, stated, This difference does not affect the applicability of TSTF-554 to the Hatch, Units 1 and 2, TSs.

Based on the above, the NRC staff finds that the licensees proposed variation from TSTF-554 are reasonable and does not affect the applicability of TSTF-554 to Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC staff, therefore, finds this variation acceptable.

The STS markup of TS 3.4.4, RCS Operational LEAKAGE, in TSTF-554 renames Required Actions B.1 and B.2 to Required Actions C.1 and C.2. STS Required Action B.2 states, Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

Hatch, Units 1 and 2, Required Action B.1 is renamed C.1, but the Hatch, Units 1 and 2, TSs do not contain an equivalent to Required Action B.2. The NRC staff finds that this difference does not affect the applicability of TSTF-554 to the Hatch, Units 1 and 2, TSs. The NRC staff, therefore, finds this variation acceptable.

3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State and Alabama State officials were notified on December 17, 2021. On December 17 and 20, 2021, the State officials of Georgia and Alabama, respectively, stated that they had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration as published in the Federal Register on November 30, 2021 (86 FR 67990), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributor: John G. Lamb, NRR/DORL Date: February 2, 2022

ML21349A518 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC OGC - NLO NAME JLamb KGoldstein VCusumano STurk DATE 12/10/2021 12/21/2021 01/13/2022 01/10/2022 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley JLamb DATE 02/02/2022 02/02/2022

v t e Letter Sequence Approval
EPID:L-2021-LLA-0175, TSTF-554 (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers Results Approval Other:
MONTHYEAR2021-09-29 [Table View]

ML21349A518

Text

SUBJECT:

Dear Ms. Gayheart:

Enclosures:

Attachment:

Attachment:

Attachment:

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RESPONSE

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Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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