ML25230A081

Units 3 and 4 — Issuance of Amendment to Revise Technical Specification 3.3.15 and 3.3.16

ML25230A081
Person / Time
Site:	Vogtle
Issue date:	09/02/2025
From:	John Lamb NRC/NRR/DORL/LPL2-1
To:	Coleman J Southern Nuclear Operating Co
References
EPID L 2025-LLA-0013 TS 3.3.15, TS 3.3.16
Download: ML25230A081 (1)

Text

September 2, 2025 Jamie Coleman Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

3535 Colonnade Parkway, Bin N-274-EC Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 ISSUANCE OF AMENDMENTS TO REVISE TECHNICAL SPECIFICATION 3.3.15 AND 3.3.16 (EPID L-2025-LLA-0013)

Dear Jamie Coleman:

In response to your application dated January 24, 2025, as supplemented by letter dated June 10, 2025, the U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 206 and 204 to Combined License (COL) Nos. NPF-91 and 92 for Vogtle Electric Generating Plant (Vogtle), Units 3 and 4, respectively. The amendments propose to move Surveillance Requirements (SRs) from Technical Specifications (TSs) 3.3.15, Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Operating, and TS 3.3.16, Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Shutdown.

A copy of the related Safety Evaluation, which includes the NRC staffs evaluation of the amendment, is enclosed. The notice of issuance of the amendment documents included in this letter will be published in the Federal Register.

If you have questions, please contact me at 301-415-0610 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.: 52-025 and 52-026

Enclosures:

1. Amendment No. 206 to Vogtle, Unit 3, COL

2. Amendment No. 204 to Vogtle, Unit 4, COL

3. Safety Evaluation cc: Listserv SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 3 DOCKET NO.52-025 AMENDMENT TO FACILITY COMBINED LICENSE Amendment No. 206 License No. NPF-91 1.

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

A.

The application for amendment by Southern Nuclear Operating Company (SNC),

dated January 24, 2025, as supplemented by letter dated June 10, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will be constructed and 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 amendment will not be inimical to the common defense and security or 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 Appendix A, Technical Specifications, of the facility Combined License as indicated in the attachment to this license amendment. Paragraph 2.D(8) of facility Combined License No. NPF-91 is hereby amended to read as follows:

(8)

Incorporation The Technical Specifications, Environmental Protection Plan, in Appendices A and B, respectively of this license, as revised through Amendment No. 206, are hereby incorporated into this license.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days of 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 Date of Issuance: September 2, 2025

Attachment:

1.

Page 4 of the facility Combined License and affected pages of Appendix A of the facility Combined License MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.09.02 13:58:46 -04'00' SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 4 DOCKET NO.52-026 AMENDMENT TO FACILITY COMBINED LICENSE Amendment No. 204 License No. NPF-92 1.

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

A.

The application for amendment by Southern Nuclear Operating Company (SNC),

dated January 24, 2025, as supplemented by letter dated June 10, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to Appendix A, Technical Specifications, of the facility Combined License as indicated in the attachment to this license amendment. Paragraph 2.D(8) of facility Combined License No. NPF-92 is hereby amended to read as follows:

(8)

Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 204, are hereby incorporated into this license.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days of 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 Date of Issuance: September 2, 2025

Attachment:

Page 4 of the facility Combined License and affected pages of Appendix C of the facility Combined License MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.09.02 13:59:24 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 206 AND 204 TO FACILITY COMBINED LICENSE NOS. NPF-91 AND NPF-92 DOCKET NOS.52-025 AND 52-026 Replace the following pages of the Facility Combined License Nos. NPF-91 and NPF-92 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Combined License No. NPF-91 REMOVE INSERT 4

4 Facility Combined License No. NPF-92 REMOVE INSERT 4

4 Appendix C to facility Combined License Nos. NPF-91 and NPF-92 REMOVE INSERT 3.3.15-2 3.3.15-2 3.3.16-3 3.3.16-3 3.3.16-4 3.4.18-1 3.4.18-2 3.4.19-1 3.4.19-2 3.4.20-1 3.5.9-1 3.5.9-2 3.7.3-1 3.7.3-1 3.7.3-2 3.7.3-2 3.7.7-1 3.7.7-1 3.7.7-2 3.7.7-2

D.

The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:

(1)

Changes during Construction - Removed by Amendment No. 202 (2)

Pre-operational Testing - Removed by Amendment Nos. 192 and 202 (3)

Nuclear Fuel Loading and Pre-critical Testing - Removed by Amendment Nos. 192 and 202 (4)

Initial Criticality and Low-Power Testing - Removed by Amendment No. 202 (5)

Power Ascension Testing - Removed by Amendment No. 202 (6)

Maximum Power Level (7)

(8)

SNC is authorized to operate the facility at steady state reactor core power levels not to exceed 3400 MW thermal (100-percent thermal power), as described in the UFSAR, in accordance with the conditions specified herein.

Reporting Requirements - Removed by Amendment No. 202 Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 206, are hereby incorporated into this license.

(9)

Technical Specifications - Removed by Amendment No. 202 (10)

Operational Program Implementation - Removed by Amendment No. 202 (11)

Operational Program Implementation Schedule - Removed by Amendment No.202 (12)

Site-and Unit-specific Conditions - Removed by Amendment No. 202

[Blank Pages 5 through 14 removed by Amendment No. 202.]

4 Amendment No. 206

D.

The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:

(1)

Changes during Construction - Removed by Amendment No. 199 (2)

Pre-operational Testing - Removed by Amendment Nos. 194 and 199 (3)

Nuclear Fuel Loading and Pre-critical Testing - Removed by Amendment Nos. 194 and 199 (4)

Initial Criticality and Low-Power Testing - Removed by Amendment No. 199 (5)

Power Ascension Testing - Removed by Amendment No. 199 (6)

Maximum Power Level (7)

(8)

SNC is authorized to operate the facility at steady state reactor core power levels not to exceed 3400 MW thermal (100-percent thermal power), as described in the UFSAR, in accordance with the conditions specified herein.

Reporting Requirements - Removed by Amendment No. 199 Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 204, are hereby incorporated into this license.

(9)

Technical Specifications - Removed by Amendment No. 199 (10)

Operational Program Implementation - Removed by Amendment No. 199 (11)

Operational Program Implementation Schedule - Removed by Amendment No. 199 (12)

Site-and Unit-specific Conditions - Removed by Amendment No. 199

[Blank Pages 5 through 14 removed by Amendment No. 199.]

4 Amendment No. 204

Technical Specifications ESFAS Actuation Logic

- Operating 3.3.15 VEGP Units 3 and 4 3.3.15 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY None

Technical Specifications ESFAS Actuation Logic

- Shutdown 3.3.16 VEGP Units 3 and 4 3.3.16 - 3 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies.

OR One or more Functions within two or more required divisions inoperable during movement of irradiated fuel assemblies.

D.1 Suspend movement of irradiated fuel assemblies.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY None

Technical Specifications Pressurizer Heater Circuit Breakers 3.4.18 VEGP Units 3 and 4 3.4.18 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Pressurizer Heater Circuit Breakers LCO 3.4.18 Each pressurizer heater circuit breaker shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, MODE 4 with all four cold leg temperatures > 275°F.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each pressurizer heater.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more pressurizer heaters with one circuit breaker inoperable.

A.1 Restore pressurizer heater circuit breaker to OPERABLE status.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.

Required Action and associated Completion Time of Condition A not met.

OR One or more pressurizer heaters with two circuit breakers inoperable.

B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND B.2 Be in MODE 4 with at least one cold leg temperature 275°F.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

Technical Specifications Pressurizer Heater Circuit Breakers 3.4.18 VEGP Units 3 and 4 3.4.18 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify each pressurizer heater circuit breaker trips open on an actual or simulated actuation signal, except for circuit breaker maintained in the tripped condition.

24 months

Technical Specifications Auxiliary Spray and Purification Line Isolation Valves 3.4.19 VEGP Units 3 and 4 3.4.19 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 Auxiliary Spray and Purification Line Isolation Valves LCO 3.4.19 Each auxiliary spray and purification line isolation valve shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS

- NOTES -

1.

Flow path(s) may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each flow path.

CONDITION REQUIRED ACTION COMPLETION TIME A.

Auxiliary spray flow path isolation valve inoperable.

OR One or two purification line flow path isolation valves inoperable.

A.1 Restore isolation valve(s) to OPERABLE status.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.

Required Action and associated Completion Time of Condition A not met.

OR Three purification line flow path isolation valves inoperable.

B.1 OR B.2 Isolate affected flow path by use of at least one closed and deactivated automatic valve.

Be In MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 6 hours

Technical Specifications Auxiliary Spray and Purification Line Isolation Valves 3.4.19 VEGP Units 3 and 4 3.4.19 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify each auxiliary spray and purification line isolation valve actuate to the isolation position on an actual or simulated actuation signal.

24 months

Technical Specifications CVS Letdown Isolation Valves 3.4.20 VEGP Units 3 and 4 3.4.20 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Chemical and Volume Control System (CVS) Letdown Isolation Valves LCO 3.4.20 Each CVS letdown isolation valve shall be OPERABLE.

APPLICABILITY:

MODE 5 below P-12 (Pressurizer Level) interlock, MODE 6 with water level < 23 feet above the top of the reactor vessel flange.

ACTIONS

- NOTE -

CVS letdown flow path may be unisolated intermittently under administrative controls.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One CVS letdown isolation valve inoperable.

A.1 Restore CVS letdown isolation valve to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B.

Required Action and associated Completion Time of Condition A is not met.

OR Two CVS letdown isolation valves inoperable.

B.1 Initiate action to isolate CVS letdown flow path by use of at least one closed and deactivated automatic valve.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.20.1 Verify CVS letdown isolation valves actuate to the isolation position on an actual or simulated actuation signal.

24 months

Technical Specifications RCP Breakers 3.5.9 VEGP Units 3 and 4 3.5.9 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.9 Reactor Coolant Pump (RCP) Breakers LCO 3.5.9 Two RCP breakers for each RCP shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4, MODE 5 with the RCS not VENTED.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each RCP.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more RCPs with one breaker inoperable in MODE 1, 2, 3 or 4.

A.1 Restore RCP breaker to OPERABLE status.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.

Required Action and associated Completion Time of Condition A is not met.

OR One or more RCPs with two breakers inoperable in MODE 1, 2, 3 or 4.

B.1 AND B.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours C.

One or more RCPs with one breaker inoperable in MODE 5.

C.1 Restore RCP breaker to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

Technical Specifications RCP Breakers 3.5.9 VEGP Units 3 and 4 3.5.9 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition C is not met.

OR One or more RCPs with two breakers inoperable in MODE 5.

D.1 Initiate action to trip affected RCP breaker(s).

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.9.1 Verify each RCP breaker trips open on an actual or simulated actuation signal, except for breaker maintained in the tripped condition.

24 months

Technical Specifications MFIVs, MFCVs, and MFW Pump Breakers 3.7.3 VEGP Units 3 and 4 3.7.3 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.7 PLANT SYSTEMS 3.7.3 Main Feedwater Isolation Valves (MFIVs), Main Feedwater Control Valves (MFCVs), and Main Feedwater (MFW) Pump Breakers LCO 3.7.3 Each MFIV, MFCV, and MFW pump breaker shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each feedwater flow path.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or both feedwater flow paths with MFIV or MFCV inoperable.

A.1 Isolate the affected flow path.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND A.2 Verify affected flow path is isolated.

Once per 7 days B.

One or both feedwater flow paths with associated MFIV and MFCV inoperable.

B.1 Isolate affected flow path.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C.

One or more feedwater flow paths with MFW pump breaker inoperable.

C.1 Restore the affected MFW pump breaker to OPERABLE status.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

Technical Specifications MFIVs, MFCVs, and MFW Pump Breakers 3.7.3 VEGP Units 3 and 4 3.7.3 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND D.2 Be in MODE 4 with the Reactor Coolant System (RCS) cooling provided by the Normal Residual Heat Removal System (RNS).

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND D.3 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1

- NOTE -

Only required to be performed prior to entry into MODE 2.

Verify the closure time of each MFIV and MFCV is within limits on an actual or simulated actuation signal.

In accordance with the Inservice Testing Program SR 3.7.3.2 Verify each MFW pump breaker trips open on an actual or simulated actuation signal, except for pump breaker maintained in the tripped condition.

24 months

Technical Specifications SFW Isolation Valves, Control Valves, and Pump Breakers 3.7.7 VEGP Units 3 and 4 3.7.7 - 1 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4) 3.7 PLANT SYSTEMS 3.7.7 Startup Feedwater (SFW) Isolation Valves, Control Valves, and Pump Breakers LCO 3.7.7 Each SFW isolation valve, control valve, and pump breaker shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

- NOTES -

1.

Flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each flow path.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more flow paths with one inoperable valve.

A.1 Isolate the affected flow path.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND A.2 Verify affected flow path is isolated.

Once per 7 days B.

One flow path with two inoperable valves.

B.1 Isolate the affected flow path.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C.

One or more flow paths with pump breaker inoperable.

C.1 Restore the affected pump breaker to OPERABLE status.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

Technical Specifications SFW Isolation Valves, Control Valves, and Pump Breakers 3.7.7 VEGP Units 3 and 4 3.7.7 - 2 Amendment No. 206 (Unit 3)

Amendment No. 204 (Unit 4)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time not met.

D.1 AND D.2 Be in MODE 3.

Be in MODE 4 with the Reactor Coolant System (RCS) cooling provided by the Normal Residual Heat Removal System (RNS).

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 24 hours AND D.3 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify each SFW isolation and control valve is OPERABLE.

In accordance with the Inservice Testing Program SR 3.7.7.2 Verify each SFW isolation and control valve actuates to the isolation position on an actual or simulated actuation signal.

24 months SR 3.7.7.3 Verify each SFW pump breaker trips open on an actual or simulated actuation signal, except for pump breaker maintained in the tripped condition.

24 months

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 206 AND 204 TO THE COMBINED LICENSE NOS. NPF-91 AND NPF-92 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 DOCKET NOS.52-025 AND 52-026

1.0 INTRODUCTION

By letter dated January 24, 2025 (Agencywide Documents Access and Management System Accession No. ML25024A238), as supplemented by letter dated June 10, 2025 (ML25161A087), Southern Nuclear Operating Company (SNC, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) amend the technical specification (TSs) for Vogtle Electric Generating Plant (Vogtle), Units 3 and 4, Combined License (COL)

Nos. NPF-91 and NPF-92, respectively. The requested license amendment request (LAR) proposed to move Surveillance Requirements (SRs) from TS 3.3.15, Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Operating, and TS 3.3.16, Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Shutdown, to other existing and new TS Limiting Conditions for Operation (LCOs).

The supplement dated June 10, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 18, 2025 (90 FR 12573).

2.0 REGULATORY EVALUATION

2.1 Background

The Protection and Safety Monitoring System (PMS) is the aggregate of electrical and mechanical equipment which senses generating station conditions and generates the signals to actuate reactor trips and engineered safety features (ESF) functions, and which provides the equipment necessary to monitor plant safety-related functions during and following designated events.

The PMS subsystems include:

Nuclear Instrumentation System (NIS),

Reactor Trip System (RTS),

Engineered Safety Features Actuation System (ESFAS), and Qualified Data Processing System (QDPS).

The PMS design criteria are the following:

Single failure does not prevent trip/actuation, even with a channel bypassed, Diversity of reactor trip functions, and Protection and control interaction o Setpoints take no credit for control actions, o Analysis assumes worst case control action/ inaction, and o Control system failures isolated from PMS.

By letter dated August 25, 2025 (ML25219A046), the NRC issued Amendment Nos. 205 and 203 for Vogtle, Units 3 and 4, respectively. Amendment Nos. 205 and 203 revised the TSs to remove the Table of Contents and allow it to be maintained by the licensee to reflect future NRC-approved amendments.

2.2 Licensee Proposed Changes Current TS, TS 3.3.15 and TS 3.3.16 provide SRs to verify that various components actuate on an actual or simulated actuation, including:

SR 3.3.15.1 for pressurizer heater circuit breakers SR 3.3.15.2 and SR 3.3.16.1 for reactor coolant pump breakers SR 3.3.15.3 for main feedwater and startup feedwater pump breakers SR 3.3.15.4 for auxiliary spray and purification line isolation valves SR 3.3.16.2 for Chemical and Volume Control System (CVS) letdown isolation valves Licensee Proposed TS Changes SNC proposed the following TS changes:

Update the TS table of contents to reflect the associated changes Create four (4) new limiting conditions for operation (LCOs) to have a location to move four (4) of the SRs:

o TS 3.4.18, Pressurizer Heater Circuit Breakers o TS 3.4.19, Auxiliary Spray and Purification Line Isolation Valves o TS 3.4.20, Chemical and Volume Control System (CVS) Letdown Isolation Valves o TS 3.5.9, Reactor Coolant Pump (RCP) Breakers Expand the scope of two (2) LCOs for two (2) of the SRs:

o TS LCO 3.7.3 would add a Main Feedwater (MFW) requirement and be revised to Main Feedwater Isolation Valves (MFIVs), Main Feedwater Control Valves (MFCVs), and MFW Pump Breakers o TS LCO 3.7.7 would add a pump breaker requirement and be revised to Startup Feedwater (SFW) Isolation Valves, Control Valves, and Pump Breakers Revise SR Tables for TS 3.3.15 and TS 3.3.16 to reflect None The proposed changes to TS 3.3.15 and TS 3.3.16 move all the SRs as follows:

o SR 3.3.15.1 is moved to new SR 3.4.18.1 o SR 3.3.15.2 and SR 3.3.16.1 are moved to new SR 3.5.9.1 o SR 3.3.15.3 is moved to new SR 3.7.3.2 and new SR 3.7.7.3 o SR 3.3.15.4 is moved to new SR 3.4.19.1 o SR 3.3.16.2 is moved to new SR 3.4.20.1 o The SR Tables for TS 3.3.15 and TS 3.3.16 are revised to reflect None Proposed Changes to Technical Specification Bases Consistent with Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.36(a)(1),

the licensee submitted corresponding changes to the TS Bases that provide the reasons for the proposed TS changes. The regulation at 10 CFR 50.36(a)(1) states that [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications. The licensee shall make changes to the Vogtle TS Bases in accordance with TS 5.5.6, Technical Specifications (TS) Bases Control Program.

2.3 Regulations The NRC staff considered the following regulatory requirements in reviewing the LAR.

The regulation at 10 CFR 50.36(c)(2), Limiting conditions for operation, states, in part, that:

(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following [four]

criteria:

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier The regulation at 10 CFR 50.36(c)(3), Surveillance requirements, states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.0 TECHNICAL EVALUATION

Under 10CFR50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis information discussed in Section 2.3 of this safety evaluation. The NRC staff reviewed the licensees submittal, as supplemented (and associated attachments and reference documents), and the relevant sections of the licensees Updated Final Safety Analysis Report (UFSAR), TS, and TS Bases to determine if the proposed changes are acceptable. The NRC staff reviewed the proposed changes to technical specifications to determine whether they meet the requirements of 10CFR50.36 and provide reasonable assurance that operation within the revised TS will not endanger the health and safety of the public.

Each SR proposed to be changed is discussed separately in the sections below.

3.1 Pressurizer Heater Circuit Breakers: SR 3.3.15.1 3.1.1 Proposed Changes The LAR proposes the creation of a new TS 3.4.18, Pressurizer Heater Circuit Breakers, and the conversion of existing SR 3.3.15.1 to new SR 3.4.18.1. The Applicability of LCO 3.4.18, and thus of SR 3.4.18.1, corresponds to the operational conditions in which existing SR 3.3.15.1 is required to be met (i.e., Modes 1, 2, and 3, and Mode 4 with all four cold leg temperatures above 275 degrees Fahrenheit (°F)).

The Conditions, Required Actions, Completion Times, and Surveillance Requirement of proposed TS 3.4.18 correspond to those already present in TS 3.3.15, with the following exceptions, each of which will be dispositioned in Section 3.1.2 of this evaluation:

Required Action B.2 in proposed LCO 3.4.18 requires being in Mode 4 with at least one cold leg temperature 275°F within a Completion Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Required Action B.2 in LCO 3.3.15 currently requires being in Mode 5 within a Completion Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Proposed SR 3.4.18.1 requires verification each pressurizer heater circuit breaker trips open on an actual or simulated actuation signal, except circuit breakers maintained in the tripped condition.

3.1.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for separate Condition entry for each pressurizer heater, for consistency with the PMS/ESFAS/pressurizer heater circuit breaker functionality as described in UFSAR Chapter 5.4.5, Pressurizer, and Chapter 7.7.1.6, Pressurizer Pressure Control System.

The protocol for performance of proposed SR 3.4.18.1 upon restoration (closure) of a circuit breaker that was excepted previously due to being maintained tripped open is acceptable based on conformance to SR 3.0.1, which requires SRs to be met during specified modes and conditions. To be declared operable, the non-tripped circuit breaker must meet the SR within the required frequency.

As described in UFSAR Sections 5.4.5 and 7.7.1.6, the pressurizer provides for pressure control of the reactor coolant system (RCS) during both steady-state operations and transient conditions by providing a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions. A surge line connecting the pressurizer to one reactor coolant hot leg allows for continuous pressure and volume adjustments between the pressurizer and the RCS. The pressurizer also serves as the initial inventory source for keeping the RCS full in the event of a small-break loss of coolant accident (LOCA).

Electrical direct-immersion heaters installed in the pressurizer bottom head keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure. As described in UFSAR Section 7.3.1.2.23 and depicted in Figure 7.2-1, the pressurizer heaters are tripped on a core makeup tank actuation signal and a High-3 pressurizer water level signal.

Division A of PMS provides actuation signals to five load center circuit breakers, which provide the power feed to five pressurizer heater electrical control centers. When these five power feed breakers are opened, the electrical power is removed from the pressurizer heaters. In addition, each division of PMS provides a separate signal to the plant control system, which are voted two-out-of-four and the result used to open the pressurizer heater circuits at the motor control center. This provides redundant breakers to trip the pressurizer heaters.

Transient and accident analyses that assume the trip of pressurizer heaters include inadvertent operation of the Core Makeup Tanks (CMTs) during power operation (UFSAR Section 15.5.1),

loss of normal feedwater flow (UFSAR Section 15.2.7; analysis bounded by the event in UFSAR Section 15.5.1), and CVS malfunctions that increase RCS inventory (UFSAR Section 15.5.2).

For these non-LOCA events, tripping the pressurizer heaters reduces pressurizer level swell.

Pressurizer heater trips also reduce the potential for steam generator overfill and automatic depressurization system actuation (i.e., Automatic Depressurization System (ADS) Stage 1, 2, and 3 actuations) during a steam generator tube rupture (SGTR) event (UFSAR Section 15.6.3),

and support RCS depressurization following CMT actuation during a small-break LOCA (UFSAR Section 15.6.5.4B). The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in the UFSAR sections described above and reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, proposed TS 3.4.18.

Based on the results of this review, the NRC staff found that creating TS 3.4.18 for pressurizer heater circuit breakers and carrying the operational conditions in which existing SR 3.3.15.1 is required to be met (i.e., Modes 1, 2, and 3, and Mode 4 with all four cold leg temperatures above 275 °F) over to the Applicability of LCO 3.4.18 are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(c)(2) and (3).

The NRC staff also compared the proposed LCO 3.4.18, Required Action B.2 requirements to those imposed by LCO addressing similar losses of pressurizer heater trip function. TS 3.3.8, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, Function 10, includes the pressurizer heater trip on Pressurizer Water Level - High 3. Pressurizer heaters are also required to trip on CMT actuation (e.g., TS 3.3.8 Function 7). Loss of one of these Functions requires placing the unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and Mode 4 with at least one cold leg temperature at or below 275°F (Function 10) or with the RCS cooling provided by the Normal Residual Heat Removal System (RNS) (Function 7) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. As stated in the applicable TS Bases, the Completion Times for these Mode change Required Actions are considered reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems, to ultimately exit the Applicability of the associated LCO. The proposed 36-hour Completion Time for Required Action B.2 of LCO 3.4.18, Condition B, is inconsistent with the 24-hour LCO 3.3.8 Completion Times to reach Mode 4 following loss of function (i.e., inability to accommodate a single failure) for ESFAS pressurizer heater trips. In its supplement dated June 10, 2025, the licensee stated that the 36-hour Completion Time to exit the applicability is an administrative change as that value is being moved from the existing action. While the 36-hour Completion Time is not consistent with the 24-hour Completion Times of similar Actions to be in MODE 4, the NRC staff find this acceptable as it is the existing Completion Time to exit the mode of applicability for Required Action B.2 of LCO 3.3.15 for the pressurizer heater circuit breakers.

In addition, SNC proposed changing the Required Action B.2 in new LCO 3.4.18 from Be in MODE 5 to Be in MODE 4 with at least one cold leg temperature 275°F. The existing LCO 3.3.15 has applicability in MODES 1, 2, 3 and 4. However, the existing SR 3.3.15.1, has a NOTE stating Only required to be met when all four cold leg temperatures are > 275°F.

Therefore, the NRC staff finds that the proposed change from Be in MODE 5 to Be in MODE 4 with at least one cold leg temperature 275°F, exits the LCO applicability, as desired.

The NRC staff considered the implications of excepting pressurizer heater circuit breakers maintained in the tripped condition from performance of SR 3.4.18.1. The transient and accident analyses in UFSAR Sections 15.2.7, 15.5.1, 15.5.2, 15.6.3, and 15.6.5 each include an automatic block of pressurizer heaters following a High-3 pressurizer level signal or CMT actuation signal. However, the initial conditions of those analyses also assume that the pressurizer heaters are functional, and the pressurizer heaters are assumed to operate during the initial progression of each event until a pressurizer heater trip signal is received.

Operability of a sufficient number of pressurizer heaters to support the initial condition assumptions in the transient and accident analyses is assured by compliance with TS 3.4.5, Pressurizer, which is applicable in the same modes/conditions as the proposed LCO 3.4.18.

Assuming compliance with TS 3.4.5, such that either sufficient pressurizer heater circuit breakers are closed or capable of closing to maintain pressurizer water level, or appropriate Required Actions are taken, the safety function of a given pressurizer heater circuit breaker relative to TS 3.4.18 is to open, consistent with the transient and accident analyses in the UFSAR. A circuit breaker that is maintained open does not require surveillance to ensure that it will open on a trip signal, as it is already in the necessary position. Consequently, excepting pressurizer heater circuit breakers that are maintained open from SR 3.4.18 does not adversely affect the safety analyses in the UFSAR and is in conformance with 10 CFR 50.36(c)(3).

3.

1.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the new TS 3.4.18, Pressurizer Heater Circuit Breakers LCO, Applicability, Actions and SRs are consistent with those in original SR 3.3.15.1, and in compliance with the regulations in 10 CFR 50.36(c)(2) and (c)(3). Therefore, NRC staff finds the relocation of SR 3.3.15.1 to TS 3.4.18 acceptable.

3.2 Reactor Coolant Pump (RCP) Breakers: SR 3.3.15.2 and SR 3.3.16.1 3.2.1 Proposed Changes The LAR proposes the creation of a new TS 3.5.9, Reactor Coolant Pump (RCP) Breakers, which includes a new SR 3.5.9.1 to replace both SR 3.3.15.2 and SR 3.3.16.1. The Applicability of LCO 3.5.9, and thus of SR 3.5.9.1, corresponds to the operational conditions in which one or more CMTs are required to be operable under TS 3.5.2, Core Makeup Tanks (CMTs) -

Operating, and TS 3.5.3, Core Makeup Tanks (CMTs) - Shutdown, Reactor Coolant System (RCS) Intact (i.e., Modes 1, 2, 3, and 4, and Mode 5 with the RCS not vented).

The Conditions, Required Actions, Completion Times, and Surveillance Requirement of proposed TS 3.5.9 correspond to those already present in TS 3.3.15 and TS 3.3.16 for the operational conditions falling within the Applicability of proposed LCO 3.5.9, with the following exceptions, each of which will be dispositioned in Section 3.2.2 of this evaluation:

Required Action D.1 in proposed LCO 3.5.9 requires the licensee to Initiate action to trip affected RCP breaker(s) with an immediate Completion Time. The Required Actions for Condition B of LCO 3.3.16, which corresponds to Condition D in proposed LCO 3.5.9, require the immediate completion of the following three Required Actions:

o Required Action B.1 to suspend positive reactivity additions.

AND o Required Action B.2 to initiate action to open RCS pressure boundary and establish 20% pressurizer level.

AND o Required Action B.3 to Initiate action to isolate the flow path from the demineralized water storage tank to the RCS by use of at least one closed and de-activated automatic valve or closed manual valve.

Proposed SR 3.5.9.1 requires the licensee to Verify each RCP breaker trips open on an actual or simulated actuation signal, except for breaker maintained in the tripped condition, on a 24 month Completion Time.

3.2.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for separate Condition entry for each RCP and Conditions defined based on the number of breakers inoperable for a given RCP, for consistency with the RCP breaker functionality as described in current existing TS Bases for 3.3.15.

The protocol for performance of proposed SR 3.5.9.1 upon restoration (closure) of a circuit breaker that was excepted previously due to being maintained tripped open is acceptable based on conformance to SR 3.0.1, which requires SRs to be met during specified modes and conditions. To be declared operable, the non-tripped circuit breaker must meet the SR within the required frequency.

UFSAR Section 5.4.1 describes the RCPs. Reactor coolant is circulated through two parallel loops, each containing a steam generator (SG) and two RCPs. The RCPs provide a core cooling flow rate that allows for sufficient heat transfer to maintain the departure from nucleate boiling ratio (DNBR) above the DNBR limit established in the safety analyses. Each RCP is provided with redundant series breakers, one from Division B and the second from Division C.

UFSAR Section 6.3 describes the passive core cooling system, including the CMTs.

Two redundant CMTs supply adequate borated water to provide RCS reactivity and inventory control during all design-basis accidents described in Chapter 15 of the UFSAR, including both LOCA and non-LOCA events.

Injection flow from the CMTs increases if the RCPs have tripped and coasted down. The transient and accident analyses summarized in UFSAR Section 6.3.3 and described in UFSAR Chapter 15 assume an RCP trip (following applicable time delay and pump coastdown) occurs whenever a CMT actuation occurs. Signals that initiate the alignment of the CMTs for injection are described in UFSAR Section 7.3.1.2.3, while conditions that result in an RCP trip are discussed in UFSAR Section 7.3.1.2.5. The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in those UFSAR sections. The staff also reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, proposed TS 3.5.9.

Based on the results of this review, the NRC staff found that creating TS 3.5.9 for RCP breakers and establishing the Applicability as the operational conditions in which the CMTs are required to be operable (i.e., Modes 1, 2, 3, 4, and Mode 5 with the RCS not vented) are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(c)(2) and (3).

The NRC staff evaluated the change from the existing Required Actions in Condition B of LCO 3.3.16 to the proposed Required Actions in Condition D of LCO 3.5.9. Both represent compensatory actions to be taken should one or more RCPs have both breakers inoperable in Mode 5 or should the Required Action and Completion Time of the Condition address the inoperability of one breaker for one or more RCPs in Mode 5 not be met. Proposed Required Action D.1 requires action be initiated to trip affected RCP breaker(s) immediately. Assuming compliance is maintained with other applicable TS (e.g., TS 3.4.4, RCS Loops), tripping the affected RCP breaker(s) restores the credited function that has been lost, eliminating the need to complete other compensatory actions. Immediately is the shortest feasible Completion Time. Additionally, if plant conditions allow, exiting the Applicability of the LCO remains an acceptable alternative per LCO 3.0.2.

The NRC staff considered the implications of excepting RCP breakers maintained in the tripped condition from performance of SR 3.5.9.1. As described above, the transient and accident analyses in UFSAR Chapter 15 assume an RCP trip (following applicable time delay and pump coastdown) occurs whenever a CMT actuation occurs. However, Chapter 15 UFSAR analyses also assume RCS loops, including RCPs, operate prior to and during the course of transients initiated from most operating conditions, in order to ensure compliance with the DNBR safety limit (SL).

Operability of RCS loops, including RCPs in operation, is assured by compliance with TS 3.4.4, RCS Loops, which is applicable in Modes 1 and 2, and Modes 3, 4, and 5 with the Plant Control System capable of rod withdrawal or one or more rods not fully inserted. Assuming compliance with TS 3.4.4, such that either sufficient RCS loops remain operable with RCPs in operation, or appropriate Required Actions are taken, the safety function of a given RCP breaker relative to TS 3.5.9 is to open, consistent with the transient and accident analyses in the UFSAR. A breaker that is maintained open does not require surveillance to ensure that it will open on a trip signal, as it is already in the necessary position. Consequently, excepting RCP breakers that are maintained open from SR 3.5.9.1 does not adversely affect the safety analyses in the UFSAR and is in conformance with 10 CFR 50.36(c)(3).

3.

2.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the new TS 3.5.9, Reactor Coolant Pump (RCP)

Breakers LCO, Applicability, Actions and SRs are consistent with those in the original SR 3.3.15.2, SR 3.3.16.1 and TS 3.5.2, Core Makeup Tanks (CMTs) - Operating, and TS 3.5.3, Core Makeup Tanks (CMTs) - Shutdown, Reactor Coolant System (RCS) Intact. In addition, the new TS 3.5.9 would continue to meet the regulations in 10 CFR 50.36(c)(2) and (c)(3).

Therefore, the NRC staff finds the relocation of SR 3.3.15.2 and SR 3.3.16.1 to TS 3.5.9 is acceptable.

3.3 Main Feedwater Pump Breakers: SR 3.3.15.3 3.3.1 Proposed Changes Existing SR 3.3.15.3 requires verifying that MFW and SFW breakers trip open on an actual or simulated actuation signal. The LAR proposes updating existing TS 3.7.3 to include MFW pump breakers and incorporating the MFW pump breaker portion of current SR 3.3.15.3 as new SR 3.7.3.2. The Applicability of LCO 3.3.15 is identical to the Applicability of both the current LCO 3.7.3 and the proposed updated LCO 3.7.3, including SR 3.7.3.2; i.e., Modes 1, 2, 3, and 4.

The Conditions, Required Actions, Completion Times, and Surveillance Requirement related to MFW pump breakers in the proposed TS 3.7.3 correspond to those already present in TS 3.3.15, with the following exceptions, each of which will be dispositioned in Section 3.3.2 of this evaluation:

Required Action D.2 in the updated LCO 3.7.3 requires, Be in Mode 4 with the Reactor Coolant System (RCS) cooling provided by the Normal Heat Removal System (RNS),

within a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A corresponding Required Action is not present in LCO 3.3.15, Condition B. However, Required Action D.2 in the updated LCO 3.7.3 is already present in the existing LCO 3.7.3 as Required Action C.2.

New SR 3.7.3.2 requires, Verify each MFW pump breaker trips open on an actual or simulated actuation signal, except for breaker maintained in the tripped condition.

3.3.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for separate Condition entry for the feedwater flow path and related Conditions defined based on the inoperability of the MFW pump breaker in one or more feedwater flow paths, for consistency with the functionality as described in current existing TS Bases for 3.3.15 and 3.7.3.

The protocol for performance of proposed SR 3.7.3.2 upon restoration (closure) of a circuit breaker that was excepted previously due to being maintained tripped open is acceptable based on conformance to SR 3.0.1, which requires SRs to be met during specified modes and conditions. To be declared operable, the non-tripped circuit breaker must meet the SR within the required frequency.

UFSAR Section 10.4.7 describes the condensate and feedwater system, including the MFW pumps, MFIVs, and MFCVs. The three MFW pumps operate in parallel to supply the secondary side of the SGs with high-pressure feedwater. The MFW pumps discharge into a single header that supplies two MFW lines, one supplying each SG. Each MFW pump is provided with one breaker, and the three MFW pump breakers are actuated from a single PMS division.

MFIVs are containment isolation valves, necessary to establish the containment boundary during any event requiring isolation of the containment One MFIV is installed in each of the two MFW lines, outside the containment and downstream of the associated MFCV. The MFIVs are relied upon to isolate MFW flow to the secondary side of the SGs following a high energy line break or excess feedwater flow event, while the MFCVs provide a second isolation during the same events. Closure of the MFIVs and/or MFCVs terminates events associated with feedwater line breaks upstream of the valves. MFIV and/or MFCV closure also mitigate the consequences of breaks in the steam or feedwater lines downstream of the valves, including limiting the mass and energy released during steam or feedwater line breaks inside containment and reducing cooldown effects during steam line breaks. As noted above, operability of the MFIVs and MFCVs is addressed under existing LCO 3.7.3.

As discussed in UFSAR Section 7.3.1.2.6 and Table 7.3-1, MFW pumps are designed to trip on the same signals that initiate MFIV closure, and for the same reasons discussed above.

Transient and accident analyses that assume the trip of the MFW pumps include FW system malfunctions that result in an increase in FW flow (UFSAR Section 15.1.2), inadvertent opening of a steam generator relief or safety valve (UFSAR Section 15.1.4), steam system piping failure (UFSAR Section 15.1.5), some FW system line breaks (UFSAR Section 15.2.8), and steam generator tube rupture (SGTR) events (UFSAR Section 15.6.3). The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in those UFSAR sections. The NRC staff also reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, the revised TS 3.7.3.

Based on the results of this review, the NRC staff found that modifying TS 3.7.3 to add MFW pump breakers and retaining the existing LCO Applicability (i.e., Modes 1, 2, 3, and 4) are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(2) and (3). With the exception of Required Action D.2 and its Completion Time, the proposed Applicability, Conditions, Required Actions, and Completion Times of the revised LCO 3.7.3 also remain consistent with those in the current LCO 3.3.15 for MFW pump breakers.

As part of the same review, the NRC staff evaluated the use of Required Action D.2 of the revised TS 3.7.3 as a compensatory action following the inoperability of a MFW pump breaker in one or more FW flow paths, followed by failure to meet the Required Action and Completion Time of Condition C. Required Action D.2 in the updated LCO 3.7.3 is already present in the existing LCO 3.7.3 as Required Action C.2, but no similar Required Action is present in LCO 3.3.15. Required Action D.1 of the revised LCO 3.7.3 requires being in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and Required Action D.3 of the revised LCO 3.7.3 requires being in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. These Required Actions and Completion Times coincide with those that would be taken following a similar inoperability under the current LCO 3.3.15, Condition B. Therefore, Required Action D.2, which requires being in Mode 4 with the RCS cooling provided by the RNS within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provides an additional step between the existing compensatory measures. This Required Action coincides with standard plant operations during a reactor shutdown (see UFSAR Section 5.4.7.1.2.1) and has a Completion Time consistent with other Required Actions necessitating plant entry into Mode 4. A 24-hour time frame for Mode 4 entry with the RCS cooling provided by RNS is considered reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems (also see Section 3.1.2 of this evaluation).

The NRC staff also considered the implications of excepting MFW pump breakers maintained in the tripped condition from performance of SR 3.7.3.2. As described above, the transient and accident analyses in UFSAR Chapter 15 assume a MFW pump trip during certain types of transients and accidents. The safety function of the MFW pump breakers is to support these analyses (i.e., to open on a valid trip signal). A breaker that is maintained open does not require surveillance to ensure that it will open on a trip signal, as it is already in the necessary position.

Consequently, excepting MFW pump breakers that are maintained open from SR 3.7.3.2 does not adversely affect the safety analyses in the UFSAR and is in conformance with 10 CFR 50.36(c)(3).

3.

3.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the changes to TS 3.7.3, MFIVs, MFCVs and MFW Pump Breakers LCO, Applicability, Actions and SRs are consistent with those in the original SR 3.3.15.3 and TS 3.7.3. In addition, the revised TS 3.7.3 would still meet the regulations in 10 CFR 50.36(c)(2) and (c)(3). Therefore, the NRC staff finds the relocation of SR 3.3.15.3 for the main feedwater pump breakers to TS 3.7.3 is acceptable.

3.4 Startup Feedwater Pump Breakers: SR 3.3.15.3 3.4.1 Proposed Changes Existing SR 3.3.15.3 requires verifying that MFW and SFW pump breakers trip open on an actual or simulated actuation signal. The LAR proposes updating existing TS 3.7.7 to include SFW pump breakers and incorporating the SFW pump breaker portion of current SR 3.3.15.3 as new SR 3.7.7.3. The Applicability of LCO 3.3.15 is identical to the Applicability of both the current LCO 3.7.7 and the proposed updated LCO 3.7.7, including SR 3.7.7.3; i.e., Modes 1, 2, 3, and 4.

The Conditions, Required Actions, Completion Times, and Surveillance Requirement related to SFW pump breakers in the proposed TS 3.7.7 correspond to those already present in TS 3.3.15, with the following exceptions, each of which will be dispositioned in Section 3.4.2 of this evaluation:

Required Action D.2 in the updated LCO 3.7.7 requires, Be in Mode 4 with the Reactor Coolant System (RCS) cooling provided by the Normal Residual Heat Removal System (RNS), within a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A corresponding Required Action is not present in LCO 3.3.15, Condition B. However, Required Action D.2 in the updated LCO 3.7.7 is already present in the existing LCO 3.7.3 as Required Action C.2.

New SR 3.7.7.3 requires, Verify each SFW pump breaker trips open on an actual or simulated actuation signal, except for pump breaker maintained in the tripped condition.

3.4.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for separate Condition entry of the SFW pump breaker in one or more flow paths, for consistency with the functionality as described in current existing TS Bases for 3.3.15 and 3.7.7.

SFW isolation valves are containment isolation valves, necessary to establish the containment boundary during any event requiring isolation of the containment. As mentioned above, operability of the SFW isolation and control valves is addressed under existing LCO 3.7.7.

As discussed in UFSAR Section 7.3.1.2.13 and Table 7.3-1, SFW pumps are designed to trip on the same signals that initiate closure of the SFW isolation and control valves, and for the same reasons discussed above. Transient and accident analyses that assume the trip of the SFW pumps include inadvertent opening of a steam generator relief or safety valve (UFSAR Section 15.1.4), steam system piping failure (UFSAR Section 15.1.5), some FW system line breaks (UFSAR Section 15.2.8), and SGTR events (UFSAR Section 15.6.3). The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in those UFSAR sections. The NRC staff also reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, the revised TS 3.7.7.

Based on the results of this review, the NRC staff found that modifying TS 3.7.7 to add SFW pump breakers and retaining the existing LCO Applicability (i.e., Modes 1, 2, 3, and 4) are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(c)(2) and (3). Except for Required Action D.2 and its Completion Time, the proposed Applicability, Conditions, Required Actions, and Completion Times of the revised LCO 3.7.7 remain consistent with those in the current LCO 3.3.15 for SFW pump breakers.

The NRC staff evaluated the use of Required Action D.2 of the revised TS 3.7.7 as a compensatory action following the inoperability of a SFW pump breaker in one or more flow paths, followed by failure to meet the Required Action and Completion Time of Condition C.

Required Action D.2 in the updated LCO 3.7.7 is already present in the existing LCO 3.7.7 as Required Action C.2, but no similar Required Action is present in LCO 3.3.15. Required Action D.1 of the revised LCO 3.7.7 requires being in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and Required Action D.3 of the revised LCO 3.7.7 requires being in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. These Required Actions and Completion Times coincide with those that would be taken following a similar inoperability under the current LCO 3.3.15, Condition B. Therefore, Required Action D.2, which requires being in Mode 4 with the RCS cooling provided by the RNS within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provides an additional step between the existing compensatory measures. This Required Action coincides with standard plant operations during a reactor shutdown (see UFSAR Section 5.4.7.1.2.1) and has a Completion Time consistent with other Required Actions necessitating plant entry into Mode 4. A 24-hour time frame for Mode 4 entry with the RCS cooling provided by RNS is considered reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems (also see Section 3.1.2 of this evaluation).

The NRC staff also considered the implications of excepting SFW pump breakers maintained in the tripped condition from performance of SR 3.7.7.3. As described above, the transient and accident analyses in UFSAR Chapter 15 assume a SFW pump trip during certain types of transients and accidents. The safety function of the SFW pump breakers is to support these analyses (i.e., to open on a valid trip signal). A breaker that is maintained open does not require surveillance to ensure that it will open on a trip signal, as it is already in the necessary position.

Consequently, excepting SFW pump breakers that are maintained open from SR 3.7.7.3 do not adversely affect the safety analyses in the UFSAR and is in conformance with 10 CFR 50.36(c)(3).

3.

4.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the changes to TS 3.7.7, SFW Isolation Valves, Control Valves, and Pump Breakers LCO, Applicability, Actions and SRs are consistent with those in the original SR 3.3.15.3 and TS 3.7.7. In addition, the revised TS 3.7.7 would still meet the regulations in 10 CFR 50.36(c)(2) and (c)(3). Therefore, the NRC staff finds the relocation of SR 3.3.15.3 for the startup feedwater pump breakers to TS 3.7.7 is acceptable.

3.5 Auxiliary Spray and Purification Line Isolation Valves: SR 3.3.15.4 3.5.1 Proposed Changes The LAR proposes the creation of a new TS 3.4.19, Auxiliary Spray and Purification Line Isolation Valves, and the conversion of existing SR 3.3.15.4 to new SR 3.4.19.1. The Applicability of LCO 3.4.19, and thus of SR 3.4.19.1, corresponds to the operational conditions in which existing SR 3.3.15.4 is required to be met (i.e., Modes 1 and 2).

The Conditions, Required Actions, Completion Times, and Surveillance Requirement of proposed TS 3.4.19 correspond to those already present in TS 3.3.15 for the Modes falling within the Applicability of LCO 3.4.19, with the following exceptions, each of which will be dispositioned in Section 3.5.2 of this evaluation:

Required Action B.1 in proposed LCO 3.4.19 requires, Isolate affected flow path by use of at least one closed and deactivated automatic valve, within a Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A corresponding Required Action is not present in LCO 3.3.15, Condition B.

The new LCO 3.4.19, Required Action B.1 provides an alternative, OR Required Action B.2, Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Note 1 of proposed LCO 3.4.19, states, Flow path(s) may be unisolated intermittently under administrative controls. Existing LCO 3.3.15 contains no such allowance.

3.5.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for each flow path and Conditions defined based on the number of auxiliary spray or purification line flow path isolation valves that are inoperable, for consistency with the functionality as described in current existing TS Bases for 3.3.15 and 3.4.19.

UFSAR Section 9.3.6 describes the CVS. CVS functions include maintaining RCS fluid purity and activity level within acceptable limits and providing pressurizer auxiliary spray water for depressurization. During power operations, fluid is continuously circulated through the normal CVS purification loop from the discharge of one of the RCPs and returns to RCP suction (i.e., under these conditions, the purification loop operates at RCS pressure, in a closed loop with the RCS). The purification line is provided with three isolation valves; two valves actuated from Division A and the third valve actuated from Division C. Pressurizer auxiliary spray is used during RCS heatups and cooldowns to add chemicals or collapse the steam bubble in the pressurizer. The single active isolation valve in the auxiliary spray line is actuated from PMS Division C.

As stated in UFSAR Section 9.3.6.1.1, consistent with the transient and accident analyses in UFSAR Chapter 15, all CVS safety functions are associated with the ability to isolate CVS lines (e.g., to provide containment isolation, to preserve the RCS pressure boundary, to isolate makeup when required). Specifically, as described in UFSAR Sections 7.3.1.2.18 and 9.3.6.7, the CVS purification and auxiliary spray lines are isolated upon receipt of Low-1 pressurizer level or manual actuation of CVS makeup isolation to preserve RCS inventory and the integrity of the RCS boundary in the event of a CVS pipe break. The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in those UFSAR sections. They also reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, proposed TS 3.4.19.

Based on the results of this review, the staff found that creating TS 3.4.19 for auxiliary spray and purification line isolation valves and carrying the Modes in which existing SR 3.3.15.4 is required to be met (i.e., Modes 1 and 2) over to the Applicability of LCO 3.4.19 are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(c)(2) and (3).

The NRC staff evaluated the use of Required Action B.1 of the proposed TS 3.4.19 as a compensatory action following the inoperability of three purification line isolation valves or following the failure to meet the Required Action and associated Completion Time of Condition A for inoperability of the auxiliary spray flow path isolation valve or 1-2 purification line isolation valves. As stated above, the CVS safety functions associated with the auxiliary spray and purification line isolation valves require the valves to support isolation capability (i.e., to automatically close and/or remain closed) when an isolation signal is present. Because the proposed Required Action B.1 initiates action to place the inoperable valve(s) in the position associated with successful completion of their safety functions and ensures that they remain in that position (unless unisolated under applicable TS provisions) and will not be adversely affected by a single active failure while they remain unable to close automatically, the staff concluded that the approach is consistent with the existing UFSAR analyses.

The NRC staff considered the implications of allowing intermittent restoration of previously isolated auxiliary spray and purification line flow paths under administrative controls, as proposed in Note 1 of LCO 3.4.19. Per Section 3 of the LAR, these administrative controls consist of stationing a dedicated operator, who remains in continuous communication with the control room, at the valve controls. The proposed administrative controls will allow for rapid isolation of the affected flow path(s) when required and will only be used intermittently during the inoperability of affected valve(s). The ability for the affected valve(s) to automatically reposition to the isolation position on an actuation signal will be restored prior to declaring the affected valve(s) operable and removing the need for the administrative controls when the affected valve(s) are opened. The allowance in Note 1 is also similar in scope and rationale to similar allowances made in other TS involving isolation of flow path(s) to comply with Required Actions (e.g., TS 3.1.9, Chemical and Volume Control System (CVS) Demineralized Water Isolation Valves and Makeup Line Isolation Valves). For those reasons, the NRC staff determined that the proposed Note 1 of LCO 3.4.19 remains consistent with 10 CFR 50.36(c)(2) and is acceptable.

3.

5.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the new TS 3.4.19, Auxiliary Spray and Purification Line Isolation Valves LCO, Applicability, Actions and SRs are consistent with those in original SR 3.3.15.4, and would still meet the regulations in 10 CFR 50.36(c)(2) and (c)(3).

Therefore, NRC staff finds the relocation of SR 3.3.15.4 to TS 3.4.19 acceptable.

3.6 CVS Letdown Isolation Valves: SR 3.3.16.2 3.6.1 Proposed Changes The LAR proposes the creation of new TS 3.4.20, Chemical and Volume Control System (CVS)

Letdown Isolation Valves, and the conversion of existing SR 3.3.16.2 to new SR 3.4.20.1. The Applicability of LCO 3.4.20, and thus of SR 3.4.20.1, corresponds to the operational conditions in which existing SR 3.3.16.2 is required to be met (i.e., Mode 5 below the P-12 [pressurizer level] interlock and Mode 6 with water level < 23 feet above the top of the reactor vessel flange).

The Conditions, Required Actions, Completion Times, and Surveillance Requirement of proposed TS 3.4.20 correspond to those already present in TS 3.3.16 for the operational conditions falling within the Applicability of LCO 3.4.20, with the following exceptions, each of which will be dispositioned in Section 3.6.2 of this evaluation:

Required Action B.1 in proposed LCO 3.4.20 requires, Initiate action to isolate CVS letdown flow path by use of at least one closed and deactivated automatic valve, with an immediate Completion Time. The Required Actions for Condition B of LCO 3.3.16, corresponds to Condition B in proposed LCO 3.4.20 for Mode 5 has the following three Required Actions with Immediate Completion Times:

o Required Action B.1 requires, Suspend positive reactivity additions, AND o Required Action B.2 requires, "Initiate action to open RCS pressure boundary and establish 20% pressurizer level.

o Required Action B.3 requires, Initiate action to isolate the flow path from the demineralized water storage tank to the RCS by use of at least one closed and de-activated automatic valve or closed manual valve.

The Required Actions for Condition C of LCO 3.3.16, which corresponds to Condition B in proposed LCO 3.4.20 for Mode 6, has the following two Required Actions with Immediate Completion Times:

o Required Action C.1 requires, Suspend positive reactivity additions, AND o Required Action C.2 requires, "Initiate action to establish water level 23 feet above the top of the reactor vessel flange.

The Note for proposed LCO 3.4.20 states, CVS letdown flow path may be unisolated intermittently under administrative controls. Existing LCO 3.3.16 contains no such allowance.

3.6.2 NRC Staff Evaluation The NRC staff evaluated the function/division nomenclature conversion to component-level nomenclature, including allowance for each flow path and Conditions defined based on the number of CVS letdown isolation valves that are inoperable, for consistency with the functionality as described in current existing TS Bases for 3.3.16 and 3.4.20.

As stated in Section 3.5.2 of this evaluation, UFSAR Section 9.3.6 describes the CVS. In addition to the purification and auxiliary spray functions described previously, CVS is also used for RCS inventory control. Increases in coolant volume (e.g., expansion during heatup from cold shutdown) are accommodated by the CVS letdown flow path to the liquid radwaste system.

CVS letdown is also used to drain the RCS to mid-loop level to permit SG draining and maintenance activities, as discussed in UFSAR Section 19E.2.1.2.4. The CVS letdown flow path is isolated by two valves; one actuated from PMS Division A and the second from PMS Division D. As stated in Table 17.4-1 of the UFSAR, the two CVS letdown isolation valves automatically close to prevent excessive reactor coolant letdown and provide containment isolation; they are important to limit offsite releases following core melt accidents.

As described in UFSAR Section 7.3.1.2.22, the CVS letdown isolation valves are designed to automatically close on a containment isolation signal. They also isolate on a Low-2 hot leg level in either of the two hot leg loops, which assists the operators when draining the RCS to mid-loop level by automatically stopping the draining process if Low-2 hot leg level is reached prior to operator termination of letdown. The Low-2 hot leg level isolation can be manually blocked above the P-12 pressurizer level interlock and is automatically unblocked below the P-12 interlock (see UFSAR Figure 7.2-1 and Tables 7.3-1 and 7.3-2).

The RCS hot leg level instruments that provide the signals for the Low-2 hot leg level isolation of CVS letdown are discussed further in UFSAR Sections 19E.2.1.2.2 and 19E.3.1.3.5. These instruments are required to be operable in shutdown modes per TS 3.3.10, Engineered Safety Feature Actuation System (ESFAS) Reactor Coolant System (RCS) Hot Leg Level Instrumentation, and actuation of the CVS letdown isolation valves upon receipt of the associated signal (or a simulated signal) is the function being tested by current SR 3.3.16.2 and proposed SR 3.4.20.1.

The NRC staff reviewed the descriptions, analyses, and underlying assumptions presented in the UFSAR sections mentioned above. The staff also reviewed other TSs with scopes and/or LCO Applicability statements interfacing with, or potentially affected by, proposed TS 3.4.20 (e.g., TS 3.3.10). Based on the results of this review, the staff found that creating TS 3.4.20 for CVS letdown isolation valves and carrying the operational conditions in which existing SR 3.3.16.2 is required to be met (i.e., Mode 5 below the P-12 interlock and Mode 6 with water level < 23 feet above the top of the reactor vessel flange) over to the Applicability of LCO 3.4.19 are consistent with the analyses presented in the UFSAR and in conformance with 10 CFR 50.36(c)(2) and (3).

The NRC staff evaluated the use of Required Action B.1 of the proposed TS 3.4.20 as a compensatory action following the inoperability of two CVS letdown isolation valves or following the failure to meet the Required Action and associated Completion Time of Condition A for inoperability of one CVS letdown isolation valve. As stated earlier in this section, the CVS safety functions associated with the CVS letdown isolation valves require the valves to support isolation capability (i.e., to automatically close and/or remain closed) when an isolation signal is present.

For the purposes of CVS isolation valve inoperability in Mode 5 below the P-12 interlock, Required Action B.1 of proposed LCO 3.4.20 is intended to replace Required Actions B.1, B.2, and B.3 of existing LCO 3.3.16, as outlined above. Section 3 of the LAR indicates that the current TS 3.3.16 Conditions broadly address the actions for inoperable ESFAS Actuation Logic. The LAR justifies Required Action B.1 of new LCO 3.4.20 by stating:

These Actions are revised to be more specifically applicable to an inoperable isolation valve where isolating the flow path accomplishes the safety function.

The Actions of TS 3.3.16 Action B do not provide compensatory actions specifically applicable to supporting maintaining RCS inventory in the event of a LOCA, which is the safety function provided by the CVS letdown isolation. As such, it provides an equivalent level of safety to restoring the valve to operable status but allows for continued efforts to restore the valve(s) to operable status without imposing unnecessary additional actions that could impact outage operations. This action is similar to that for other CVS isolation valves (refer to TS 3.1.9, CVS Demineralized Water Isolation Valves and Makeup Line Isolation Valves). These changes do not adversely impact a safety function assumed in the safety analyses.

Required Action B.1 of proposed LCO 3.4.20 initiates action to place the inoperable valve(s) in the position associated with successful completion of their safety functions, and ensures that, following closure, they will remain in that position (unless unisolated under applicable TS provisions) and will not be adversely affected by a single active failure while they remain unable to close automatically. Function 2 of TS 3.3.10, Engineered Safety Feature Actuation System (ESFAS) Reactor Coolant System (RCS) Hot Leg Level Instrumentation, covers the hot leg level instruments that sense Hot Leg Level - Low 2 conditions and initiate the closure of the CVS letdown isolation valves. Isolation of the CVS letdown flow path and ensuring that the flow path remains closed while in Mode 5 below the P-12 interlock are consistent with one allowable set of remedial actions specified for extended loss of instrument function (i.e., inability to accommodate a single failure) in Condition D of LCO 3.3.10, which would have the same plant effect as an extended inoperability (inability to close) of both CVS letdown isolation valves. As a result of those considerations, the NRC staff concluded that replacement of Required Actions B.1, B.2, and B.3 of existing LCO 3.3.16, as they apply to CVS letdown isolation valve inoperability in Mode 5 below the P-12 interlock, with Required Action B.1 of proposed LCO 3.4.20 provides adequate compensatory measures for loss of the automatic protection function under those operating conditions.

As mentioned in the preceding section of this safety evaluation, for the purposes of CVS isolation valve inoperability in Mode 6 with water level < 23 feet above the top of the reactor vessel flange, Required Action B.1 of proposed LCO 3.4.20 is intended to replace Required Actions C.1 and C.2 of existing LCO 3.3.16. LCO 3.3.16 Required Action C.1 requires immediate suspension of positive reactivity additions. LCO 3.3.16 Required Action C.2 requires immediately initiating action to establish water level 23 feet above the top of the reactor vessel flange. The justification for this change from Section 3 of the LAR appears above in part.

While the Actions of Condition B in LCO 3.3.16 are not specifically relevant to maintaining RCS inventory during a LOCA or excessive reactor coolant letdown during Mode 5, Required Action C.2 of LCO 3.3.16 is directly applicable to mitigation of such events during Mode 6. While the proposed Required Action B.1 of LCO 3.4.20 is similar to Required Action B.1 of LCO 3.1.9, as mentioned in the LAR, LCO 3.1.9 is not applicable in Mode 6. TS 3.3.10, Function 2, is applicable in Mode 5 below the P-12 pressurizer level interlock and Mode 6 with water level

< 23 feet above the top of the reactor vessel flange, and extended loss of instrument function (i.e., inability to accommodate a single failure) would have the same plant effect as an extended inoperability (inability to close) of both CVS letdown isolation valves. In Mode 6 with water level

< 23 feet above the top of the reactor vessel flange, Required Action E.1 of LCO 3.3.10 requires the same Required Action (initiate action to establish water level 23 feet above the top of the reactor vessel flange) with the same Completion Time (immediately) as existing LCO 3.3.16, Required Action C.2.

In the letter dated June 10, 2025, SNC stated that it has submitted a separate LAR, which is currently under review, that proposes to replace both the TS 3.3.16 Required Action C.2 and TS 3.3.10 Required Action E.1 that require initiating action to establish water level 23 feet above the top of the reactor vessel flange. The licensee also stated that removing the prescriptive requirement (to establish a certain water level) that would exit the proposed TS 3.4.20 Applicability, does not change the option to raise level and exit the Applicability. As stated in the TS Bases for TS 3.3.16, Actions C.1 and C.2, the action to establish reactor cavity water level 23 feet above the top of the reactor vessel flange is to minimize the consequences of an event. Based on its review of the above information, the NRC staff finds that establishment of a certain water level is not a requirement and that the proposed actions in TS 3.4.20 are appropriate and consistent with other TS Actions for inoperable functions allowing continued operation with an isolated flow path providing the safety function.

The NRC staff considered the implications of allowing intermittent restoration of a previously isolated CVS letdown flow path under administrative controls, as proposed by the Note in LCO 3.4.20. Section 3 of the LAR justifies this Note by stating:

Also similar to the TS 3.1.9 actions allowing continued operation with an isolated flow path providing the safety function, the proposed TS 3.4.20 Actions includes a Note allowing the CVS letdown flow path to be unisolated intermittently under administrative control. These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way the flow path can be rapidly isolated when a need for CVS letdown line isolation is indicated. As such, this change does not adversely impact a safety function assumed in the safety analyses.

The proposed administrative controls will allow for rapid isolation of the flow path when required and will only be used intermittently during the inoperability of the affected valves. The ability for the affected valve(s) to automatically reposition to the isolation position on an actuation signal will be restored prior to declaring the affected valve(s) operable and removing the need for the administrative controls when the affected valve(s) are opened. The allowance in the Note is also similar in scope and rationale to similar allowances made in other TS involving isolation of flow path(s) to comply with Required Actions in Mode 5. These include TS 3.1.9, Chemical and Volume Control System (CVS) Demineralized Water Isolation Valves and Makeup Line Isolation Valves, and Function 2 of TS 3.3.10, Engineered Safety Feature Actuation System (ESFAS)

Reactor Coolant System (RCS) Hot Leg Level Instrumentation. As stated earlier in this section, TS 3.3.10 Function 2 covers the hot leg level instruments that sense Hot Leg Level - Low 2 conditions and initiate the closure of the CVS letdown isolation valves. In Mode 5 below the P-12 interlock, the LCO 3.3.10 Required Actions associated with flow path isolation for an extended loss of function (i.e., inability to accommodate a single failure; Required Actions D.1.1, D.1.2.1, and D.1.2.2) are modified with a Note allowing the flow path to be unisolated intermittently under administrative controls. In its letter dated June 10, 2025, SNC stated that the allowance for Mode 5 is equally applicable to Mode 6 to provide sufficient compensatory measures that the flow path can be rapidly isolated when a need for flow path isolation is indicated. As a result of those considerations, the NRC staff concluded that the proposed LCO 3.4.20 Note, as it applies to inoperable CVS letdown isolation valve(s) in Mode 5 below the P-12 interlock and in Mode 6 with water level < 23 feet above the top of the reactor vessel flange, supports adequate compensatory measures for loss of the automatic protection function under those operating conditions.

3.

6.3 NRC Staff Conclusion

Based on the above, the NRC staff concludes the new TS 3.4.20, CVS Letdown Isolation Valves LCO, Applicability, Actions and SRs are consistent with those in original SR 3.3.16.2, and would continue to meet the regulations in 10 CFR 50.36(c)(2) and (c)(3). In addition, the staff finds it acceptable to include only actions specific to the CVS letdown isolation valves that perform the safety function. This results in new TS 3.4.20 not containing Required Actions B and C from TS 3.3.16 (e.g., suspend positive reactivity and establish a specific water level);

however, these actions will remain in TS 3.3.16. Therefore, the NRC staff finds the relocation of SR 3.3.16.2 to TS 3.4.20 acceptable.

3.7 SR Tables for TS 3.3.15 and TS 3.3.16 3.7.1 Proposed Changes The proposed TS changes dispositioned above remove all existing SRs from TS 3.3.15 and TS 3.3.16. As a result, the LAR proposes to update the SR tables for those two TS to reflect None, consistent with the precedent in ML19297C791, Vogtle Electric Generating Plant, Units 3 and 4 Issuance of Amendments regarding Request for License Amendment regarding Protection and Safety Monitoring System Surveillance Requirement Reduction Technical Specification Revision (EPID L-2019-LLA-0064).

3.7.2 NRC Staff Evaluation The NRC staff previously found (see ML19297C791) that, excepting the end device actuation tests described in previous sections of this evaluation, the performance of ESF Coincidence Logic and ESF Actuation Logic SRs under TS 3.3.15 and TS 3.3.16 is not required due to the ability of PMS self-diagnostic functions to adequately demonstrate operability of the applicable components. The NRC staff evaluation in ML19297C791 also confirmed the acceptability of having no SRs for a given LCO, provided SR(s) are not required to demonstrate operability under that LCO (e.g., LCO 3.3.6, Reactor Trip System (RTS) Automatic Trip Logic). The NRC staff reviewed ML19297C791 as it relates to leaving no SRs associated with TS 3.3.15 and TS 3.3.16.

3.

7.3 NRC Staff Conclusion

Because the end device actuation tests will be moved to other new or existing TS, as evaluated above, and additional SRs to verify the operability of ESF Coincidence Logic and ESF Actuation under TS 3.3.15 and TS 3.3.16 were previously found to be unnecessary by the NRC staff in ML19297C791, the NRC staff finds that having SRs for TS 3.3.15 and TS 3.3.16 are not required to meet 10 CFR 50.36(c)(3). Thus, the NRC staff finds that having no SRs for TS 3.3.15 and TS 3.3.16, as proposed in the LAR, is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Georgia State official was notified of the proposed issuance of the amendments on August 7, 2025. On August 7, 2025, the State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and SRs. 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on March 18, 2025 (90 FR 12573). 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.

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

Principal Contributors: C. Szumski, NRR R. Beaton, NRR K. West, NRR B. Lee, NRR K. James, NRR L. Kwan, NRR N. Carte, NRR B. Rothberg, NRR Date: September 2, 2025

ML25230A081 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NRR/DSS/SCPB/BC NAME JLamb KZeleznock SMehta (ARussell for)

MValentin (BLee for)

DATE 08/16/2025 08/20/2025 08/22/2025 08/25/2025 OFFICE NRR/DSS/SNSB/BC NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME NDiFrancesco MMarkley JLamb DATE 08/22/2025 09/02/2025 09/02/2025