ML25199A048

Issuance of Expedited Amendment to Revise Technical Specification 3.3.8, Engineered Safety Feature Actuation System (ESFAS) Instrumentation

ML25199A048
Person / Time
Site:	Vogtle
Issue date:	08/08/2025
From:	John Lamb NRC/NRR/DORL/LPL2-1
To:	Coleman J Southern Nuclear Operating Co
References
EPID L 2025-LLA-0035 TS 3.3.8
Download: ML25199A048 (1)

Text

August 8, 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 EXPEDITED AMENDMENTS TO REVISE TECHNICAL SPECIFICATION 3.3.8, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM (ESFAS)

INSTRUMENTATION (EPID NO. L-2025-LLA-0035)

Dear Ms. Coleman:

In response to your letter dated February 14, 2025, as supplemented by letters dated March 31, July 9, and July 16, 2025, the U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 203 and 200 to Combined License (COL) Nos. NPF-91 and 92 for Vogtle Electric Generating Plant (Vogtle), Units 3 and 4, respectively. The amendments revise Technical Specifications (TS) 3.3.8, Engineered Safety Feature Actuation System (ESFAS)

Instrumentation.

The amendments add a new Tcold - High, Function 11.b for passive residual heat removal actuation logic, to TS Table 3.3.8-1, Engineered Safeguards Actuation System Instrumentation, as well as editorial changes to TS 3.3.8 at Vogtle, Units 3 and 4.

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.

J. Coleman If you have questions, please 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.: 52-025 and 52-026

Enclosures:

1. Amendment No. 203 to Vogtle, Unit 3, COL

2. Amendment No. 200 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. 203 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 February 14, 2025, as supplemented by letters dated March 31, July 9, and July 16, 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.

J. Coleman 2.

Accordingly, the license is amended to authorize changes to the Updated Final Safety Analysis Report (UFSAR) as described in the licensees application dated February 14, 2025, as supplemented by letters dated March 31, July 9, and July 16, 2025. The license is also amended by changes to Appendix A, Technical Specifications, of the facility Combined License as indicated in the attachment to this license amendment.

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 and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 203, are hereby incorporated into this license.

3.

This license amendment is effective as of the date of its issuance and shall be implemented prior to startup from the Spring 2026 refueling outage.

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: August 8, 2025

Attachment:

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: 2025.08.08 07:13:26 -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. 200 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 February 14, 2025, as supplemented by letters dated March 31, July 9, and July 16, 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 to authorize changes to the Updated Final Safety Analysis Report (UFSAR) as described in the licensees application dated February 14, 2025, as supplemented by letters dated March 31, July 9, and July 16, 2025. The license is also amended by changes to Appendix A, Technical Specifications, of the facility Combined License as indicated in the attachment to this license amendment.

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 and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 200, are hereby incorporated into this license.

3.

This license amendment is effective as of the date of its issuance and shall be implemented prior to startup from the Spring 2027 refueling outage.

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: August 8, 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: 2025.08.08 07:14:07 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 203 AND 200 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 3.3.8-7 3.3.8-7 3.3.8-8 3.3.8-8 3.3.8-9

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. 203, 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. 203

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. 200, 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. 200

Technical Specifications ESFAS Instrumentation 3.3.8 VEGP Units 3 and 4 3.3.8 - 7 Amendment No. 203 (Unit 3)

Amendment No. 200 (Unit 4)

Table 3.3.8-1 (page 1 of 3)

Engineered Safeguards Actuation System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS 1.

Containment Pressure

a. - Low 1,2,3,4,5(a),6(a) 4 P

b.

- Low 2 1,2,3,4,5(a),6(a) 4 P

2.

Containment Pressure - High 2 1,2,3,4 4

H 3.

Containment Radioactivity - High 1,2,3,4(b) 4 l

4.

Containment Radioactivity - High 2 1,2,3 4

l 5.

Pressurizer Pressure - Low 3 1,2,3(c)(l) 4 E

6.

Pressurizer Water Level - Low 1,2 4

D 7.

Pressurizer Water Level - Low 2 1,2,3,4(b) 4 F

4(d),5(e) 4 J

8.

Pressurizer Water Level - High 1,2,3 4

l 9.

Pressurizer Water Level - High 2 1,2,3,4(f) 4 l

10.

Pressurizer Water Level - High 3 1,2,3,4(f) 4 Q

11.

RCS Cold Leg Temperature (Tcold)

a. - Low 2 1,2,3(c)(l) 4 per loop E

b. - High 1,2,3,4(b) 4 per loop F

12.

Reactor Coolant Average Temperature (Tavg)

- Low 1,2 4

D 13.

Reactor Coolant Average Temperature (Tavg)

- Low 2 1,2 4

D (a) Without an open containment air flow path 6 inches in diameter.

(b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS).

(c) Above the P-11 (Pressurizer Pressure) interlock.

(d) With the RCS being cooled by the RNS.

(e) With RCS not VENTED and CMT actuation on Pressurizer Water Level - Low 2 not blocked.

(f)

With all four cold leg temperatures > 275°F.

(l)

Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SDM requirements at an RCS temperature of 200°F.

Technical Specifications ESFAS Instrumentation 3.3.8 VEGP Units 3 and 4 3.3.8 - 8 Amendment No. 203 (Unit 3)

Amendment No. 200 (Unit 4)

Table 3.3.8-1 (page 2 of 3)

Engineered Safeguards Actuation System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS 14.

RCS Wide Range Pressure - Low 1,2,3,4 4

H 5

4 K

6(g) 4 L

15.

Core Makeup Tank (CMT) Level - Low 3 1,2,3,4(b) 4 per tank F

4(d),5(h) 4 per OPERABLE tank J

16.

CMT Level - Low 6 1,2,3,4(b) 4 per tank F

4(d),5(h) 4 per OPERABLE tank J

17.

Source Range Neutron Flux Doubling 2(i),3(i),4(j) 4 I

5(j) 4 I

18.

IRWST Lower Narrow Range Level - Low 3 1,2,3,4(b) 4 F

4(d),5 4

M 6(g) 4 N

19.

Reactor Coolant Pump Bearing Water Temperature - High 2 1,2,3,4 4 per RCP O

20.

SG Narrow Range Water Level - Low 2 1,2,3,4(b) 4 per SG F

21.

SG Wide Range Water Level - Low 2 1,2,3,4(b) 4 per SG F

22.

SG Narrow Range Water Level High 1,2,3,4 4 per SG I

23.

SG Narrow Range Water Level - High 3 1,2 4 per SG D

3,4 4 per SG I

(b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS).

(d) With the RCS being cooled by the RNS.

(g) With upper internals in place.

(h) With RCS not VENTED.

(i)

With unborated water source flow paths not isolated except when critical or except during intentional approach to criticality.

(j)

With unborated water source flow paths not isolated.

Technical Specifications ESFAS Instrumentation 3.3.8 VEGP Units 3 and 4 3.3.8 - 9 Amendment No. 203 (Unit 3)

Amendment No. 200 (Unit 4)

Table 3.3.8-1 (page 3 of 3)

Engineered Safeguards Actuation System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS 24.

Steam Line Pressure - Low 2 1,2,3(c)(l)(m) 4 per steam line G

25.

Steam Line Pressure - Negative Rate - High 3(k) 4 per steam line I

(c) Above the P-11 (Pressurizer Pressure) interlock.

(k) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is blocked.

(l)

Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SDM requirements at an RCS temperature of 200°F.

(m) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is not blocked.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 203 AND 200 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 February 14, 2025 (Agencywide Documents Access and Management System Accession No. ML25045A166), as supplemented by letters dated March 31, July 9, and July 16, 2025 (ML25090A283, ML25190A653, and ML25197A672, respectively), Southern Nuclear Operating Company (SNC, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC) amend Vogtle Electric Generating Plant (Vogtle), Units 3 and 4, Combined Operating License (COL) Numbers NPF-91 and NPF-92, respectively. The license amendment request (LAR) proposed to revise Technical Specifications (TS) 3.3.8, Engineered Safeguards Actuation System Instrumentation.

The amendments add a new Function 11.b, Reactor Coolant System (RCS) Cold Leg Temperature (Tcold) - High to TS Table 3.3.8-1, Engineered Safeguards Actuation System Instrumentation. The new Function 11.b impacts the Protection and Safety Monitoring System (PMS) actuation logic for the Passive Residual Heat Removal (PRHR) Heat Exchanger (HX).

The logic for one of the PRHR HX actuation signals that currently requires Low-2 steam generator (SG) narrow range (NR) water level coincident with Low-2 startup feedwater (SFW) flow in any SG is being modified to require Low-2 SG NR water level coincident with Low-2 SFW flow in both SGs coincident with the new Tcold - High in either RCS loop.

The supplements dated July 9, 2025, and July 16, 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 May 14, 2025 (90 FR 20519).

2.0 REGULATORY EVALUATION

2.1 Background

During its pre-submittal meeting on October 24, 2024 (ML24290A158), SNC stated that the reason for the proposed change is to address the Haiyang Nuclear Power Plant (Haiyang), Unit 2, event that occurred on October 17, 2018. Haiyang, Unit 2, feedwater pump tripped that led to a reactor trip, and a subsequent SFW initiation; this resulted in a decrease in the cold leg temperatures. The Haiyang, Unit 2, operators attempted to prevent a safeguards signal from being generated on Tcold - Low-2. This resulted in Low-2 SG NR level coupled with Low-2 SFW flow, which generated an automatic PRHR HX actuation, reducing Tcold further leading to a safeguards actuation.

Vogtle, Units 3 and 4, TS 3.3.1 is Reactor Trip System (RTS) Instrumentation. Limiting Condition for Operation (LCO) 3.3.1 states:

The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

Vogtle, Units 3 and 4, TS 3.3.11 is Engineered Safety Feature Actuation System (ESFAS)

Startup Feedwater Flow Instrumentation. LCO 3.3.11 states:

Two channels of ESFAS Startup Feedwater Flow - Low 2 instrumentation for each startup feedwater line shall be OPERABLE.

Vogtle, Units 3 and 4, TS 3.3.8 is Engineered Safety Feature Actuation System (ESFAS)

Instrumentation. LCO 3.3.8 states:

The ESFAS instrumentation channels for each Function in Table 3.3.8-1 shall be OPERABLE.

TS Table 3.3.8-1 has, in part, the following functions:

Number Function 11 RCS Cold Leg Temperature (Tcold) - Low 2 20 SG Narrow Range Water Level - Low 2 21 SG Wide Range Water Level - Low 2 The current HX actuation logic is coincidence of 2-of-4 channels below SG NR Water Level -

Low-2 in either SG, after a preset time delay, coincident with 1-of-2 SFW flow - Low-2 feeding same SG. The current PRHR HX actuation logic is coincidence of 2-of-4 channels below SG Wide Range Water Level Low-2 in either SG.

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

The regulation at Title 10 of the Code of Federal Regulations (10 CFR), section 50.36, Technical specifications, provides the regulatory requirements for the content of the TS. It requires, in part, that a summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall be included in the application, but shall not become part of the technical specifications. Specifically, 10 CFR 50.36(c) requires that TS include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) Surveillance requirements; (4) design features; and (5) administrative controls.

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 criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(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.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulation at 10 CFR 50.55a(h), Protection and safety systems, states, in part, that:

(2) Protection systems. For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements in IEEE Std 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or the requirements in IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, or the requirements in IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations,[] and the correction sheet dated January 30, 1995.

For nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.

(3) Safety systems. Applications filed on or after May 13, 1999, for construction permits and operating licenses under this part, and for design approvals, design certifications, and combined licenses under part 52 of this chapter, must meet the requirements for safety systems in IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.

The regulation at Appendix A to Part 50, General Design Criteria [GDC] for Nuclear Power Plants, Criterion 13Instrumentation and control, states:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

The regulation at Appendix A to Part 50, Criterion 20Protection system functions, states:

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The regulation at Appendix A to Part 50, Criterion 34Residual heat removal, states:

A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Applicable criteria in Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, including:

Criterion III, Design Control, which states, in part, that:

Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which [Appendix B to 10 CFR Part 50] applies are correctly translated into specifications, drawings, procedures, and instructions.

The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

The verifying or checking process shall be performed by individuals or groups other than those who performed the original design, but who may be from the same organization.

Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.

Criterion V, Instructions, Procedures, and Drawings, states that:

Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Criterion VII, Control of Purchased Material, Equipment, and Services, states, in part, that:

Measures shall be established to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures shall include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products upon delivery.

Documentary evidence that material and equipment conform to the procurement requirements shall be available at the nuclear power plant site prior to installation or use of such material and equipment...

Criterion XI, Test Control, states in, part, that:

A test program be established to assure that testing required to demonstrate that structures, systems and components will perform satisfactorily in service in identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents...

Criterion XVI, Corrective Action, states that:

Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

3.0 TECHNICAL EVALUATION

3.1 Editorial Changes SNC proposes making several editorial changes, such as:

Adding an (a) to the RCS Cold Leg Temperature (Tcold) - Low 2 of Function 11 In the header on TS page 3.3.8-7, (page 1 of 2) becomes (page 1 of 3)

Functions 24 and 25 roll over from TS page 3.3.8-8 to a new page, TS page 3.3.8-9 Footnotes (c), (k), (l), and (m) move from TS page 3.3.8-8 to TS page 3.3.8-9 In the header on TS page 3.3.8-8, (page 2 of 2) becomes (page 2 of 3)

Add a new TS page 3.3.8-9 with a header, with moved Functions 24 and 25 and the appropriate footnotes The proposed editorial changes do not alter the intent, scope, or requirements in the TS, but do serve to improve its presentation and usability. Therefore, the NRC staff finds the proposed editorial changes acceptable.

3.2 Overview of Proposed Technical Changes The AP1000 instrumentation and control (I&C) system includes functions to sense plant process conditions and actuate engineered safety features (ESF) accordingly. The PMS is provided as part of the AP1000 I&C system to initiate ESF actuations when plant process conditions reach pre-determined setpoints. Once associated logic conditions inside the PMS are met, the PMS produces the command signals to actuate appropriate ESF components.

SNC proposed to add an RCS Cold Leg Temperature (Tcold) - High to TS Table 3.3.8-1. The proposed PRHR HX actuation logic is provided by the coincidence of two of the four divisions of SG NR water level below the Low-2 setpoint after a preset time delay, a coincident Low-2 SFW flow in both SGs, and coincident with two of the four divisions of Tcold - High.

SNC stated that the addition of the new Tcold - High does not require any physical plant modifications. The RCS Tcold hardware and temperature input into PMS currently exists as required by TS Table 3.3.8-1, Function 11, Tcold - Low-2. The modification to implement a new Tcold - High Function will utilize the existing safety-related PMS resistance temperature detectors for RCS Tcold.

3.3 Current TSs The current Function 11 of Table TS 3.3.8-1 states:

FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS

11. RCS Cold Leg Temperature (Tcold) - Low 2 1,2,3(c)(l) 4 per loop E

3.4 Proposed TSs The proposed Function 11 of Table TS 3.3.8-1 would state:

FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS

11. RCS Cold Leg Temperature (Tcold)

a. - Low 2 1,2,3(c)(l) 4 per loop E

b. - High 1,2,3,4(b) 4 per loop F

3.5 Technical Evaluation of the Proposed Changes In Section 3. of its submittal dated February 14, 2025, the licensee stated that the following events could be affected by the requested change because PRHR actuation is credited for accident mitigation:

The loss of alternating current (LOAC) accident as described in UFSAR subsection 15.2.6,

The loss of normal feedwater flow (LONF) accident as described in UFSAR subsection 15.2.7, and

The LONF with LOAC accident as described in UFSAR subsection 15.2.7 3.5.1 Technical Review of Loss of Alternating Current Event UFSAR subsection 15.2.6 describes plant response to a LOAC to the plant auxiliaries, which is caused by a complete loss of the offsite grid and a turbine-generator trip. Although the onsite AC system remains available, it is not credited for accident mitigation. In current plant operations, under a LOAC scenario, the startup feedwater system removes core decay heat if it is available. If startup feedwater is unavailable, the PRHR heat exchanger will provide emergency core decay heat removal, actuated by the Low-2 SG NR water level coincident with Low-2 start up feedwater flow.

The proposed change would alter this response by requiring Low-2 SG NR level coincident with Low-2 start up feedwater flow on both SGs and coincident with two of the four divisions of Tcold -

High. This avoids actuation of the PRHR system if there is an intact SG available as a heat sink.

The licensee analyzed the LOAC event with the proposed changes using the LOFTRAN computer code and determined that it remained within acceptance criteria which includes pressurizer volume maximum value and minimum departure from nucleate boiling ratio (DNBR).

The NRC staff reviewed the LOAC event sequence in UFSAR Table 15.2-1 and concludes that the accident evaluation remains valid because the initial conditions assumed a time delay once the Low-2 SG NR setpoint is reached. The PRHR HX will continue to actuate at 637.0 seconds, as described in UFSAR Table 15.2-1 (sheet 5 of 8), indicating there will be no change to the accident analysis for LOAC resulting from this proposed change.

3.5.2 Review of Loss of Normal Feedwater Flow UFSAR subsection 15.2.7 describes plant response to a LONF which results in a reduction in the capability of the secondary system to remove decay heat due to a loss of normal feedwater from pump failures, valve malfunctions, or loss of ac power sources. Following this event, the PRHR heat exchanger removes long-term decay heat, actuated on either a Low-2 SG NR (narrow range) water level, coincident with a Low-2 startup feedwater flow rate signal or a Low-2 SG WR (wide range) water level signal.

The proposed change would alter this response by requiring Low-2 SG NR water level coincident with Low-2 start up feedwater flow on both SGs and coincident with two of the four divisions of Tcold - High. This avoids actuation of the PRHR system if there is an intact SG available as a heat sink. The SG WR level Low-2 is unaffected by the proposed change. The licensee analyzed the LONF event with the proposed changes using the LOFTRAN computer code and determined that it remained within acceptance criteria which includes pressurizer volume maximum value and minimum DNBR. The NRC staff reviewed the LONF event sequence in UFSAR Table 15.2-1 and concluded that the accident evaluation remains valid because the initial conditions assumed a time delay once the Low-2 SG NR setpoint is reached.

The PRHR HX will continue to be actuated at 243.1 seconds, as described in UFSAR Table 15.2-1 (sheet 6 of 8), indicating there will be no change to the accident analysis for LONF resulting from this proposed change.

The combined LONF with LOAC event (UFSAR 15.2.7) has different initial conditions from the LONF event because losing AC power results in a coast down of the reactor coolant pumps instead of maintaining them at normal speed until the automatic trip from the safeguards actuation (S) signal. This difference between reactor coolant natural circulation and forced flow affects when systems reach their setpoints which affects the event timing. The licensee determined the proposed change would not affect the existing accident mitigation strategy because the PRHR heat exchanger is not actuated on the SG NR level. NRC staff reviewed the LONF with LOAC event sequence in UFSAR Table 15.2-1 and confirmed the PRHR heat exchanger is actuated on the SG WR level Low-2 setpoint which is unaffected by the proposed change and that the accident evaluation remains valid.

3.5.3 Review of I&C Logic The proposed logic change requires:

Both SGs to reach Low-2 narrow range level and Low-2 startup feedwater flow.

The addition of a new Reactor Coolant System (RCS) Tcold - High function.

The proposed logic requires both SGs to reach Low-2 levels, ensuring that PRHR actuation only occurs when both SGs are compromised, thus avoiding unnecessary actuation and maintaining the availability of the heat sink. The new RCS Tcold - High actuation is set below the turbine bypass valve pressure control mode setpoint. This anticipatory actuation prevents the SG power-operated relief valves or turbine bypass valves from depleting the SG inventory if startup feedwater flow is lost. This ensures that the PRHR HX is actuated in a timely manner to maintain core cooling and prevent overheating, enhancing the overall safety and reliability of the system.

The logic change maintains the single-failure proof design by ensuring redundancy in signal generation:

SG narrow range Low-2 signal is generated from 2 out of 4 divisions for each SG.

Startup feedwater flow Low-2 signal is generated from 1 out of 2 divisions for each SG.

RCS Tcold - High signal is generated from 2 out of 4 divisions for either RCS loop.

This redundancy ensures that the system remains robust and reliable even in the event of a single component failure.

The proposed logic change for the PRHR HX actuation enhances the safety and reliability of the system by ensuring that actuation occurs only when both SGs are compromised, preventing unnecessary actuation, and maintaining a single-failure proof design. The addition of the RCS Tcold - High function provides anticipatory protection, further enhancing the system's safety performance.

On July 21, 2025, the NRC staff performed a regulatory audit at Westinghouse Electric Corporation (WEC) in Warrendale, Pennsylvania. During this audit, the NRC staff traced a sample of system requirements for the plant protection system logic change from the design phase through to testing. For these traced requirements, the WEC staff was able to demonstrate both forward and backwards traceability. Although the testing phase had not started at the time of the audit, the NRC staff observed that the tracing requirements included applicable test procedures. The NRC staff documented its observations in the audit summary (ML25209A442).

3.5.4 Review PRHR HX Alignment The PRHR HX is aligned via a natural circulation loop connecting the RCS hot leg and cold leg through a heat exchanger submerged in the In-containment Refueling Water Storage Tank.

Upon actuation, this configuration enables passive heat removal without reliance on active systems or operator actions, enhancing plant safety under station blackout or loss-of-feedwater conditions.

Currently, PRHR HX actuates when one out of two SGs reaches Low-2 narrow range water level for a predefined time delay, coincident with one-out-of-two SGs at Low-2 startup feedwater flow. The proposed modification revises PRHR HX activation criteria, requiring both SGs to reach Low-2 narrow range level and Low-2 startup feedwater flow. This change prevents premature PRHR activation caused by inventory loss in only one SG while the other remains available as a heat sink. Additionally, a new RCS Tcold - High function has been introduced, to ensure activation before SG inventory depletion via turbine bypass valves or power-operated relief valves.

The NRC staff evaluated the impact of incorporating the new RCS Tcold - High signal as a coincidence condition in the PRHR HX actuation logic and found that the revised initiation logic would actuate when the following conditions are simultaneously met: Low-2 SG NR water level coincident with Low-2 startup feedwater flow coincident with High cold leg temperature (Tcold -

High).

3.5.5 Review of Proposed TS 3.3.8 Change SNC proposed to add Function 11.b, RCS Cold Leg Temperature (Tcold) - High to TS Table 3.3.8-1. The proposed PRHR HX actuation logic is provided by the coincidence of two of the four divisions of SG NR water level below the Low-2 setpoint after a preset time delay, a coincident Low-2 SFW flow in both SGs, and coincident with two of the four divisions of Tcold -

High.

Four channels are provided to permit one channel to be in trip or bypass condition indefinitely and still ensure no single random failure will disable this Function. The setpoint reflects both steady state and adverse environmental instrument uncertainties, as the detectors provide protection for an event that results in a harsh environment. PRHR HX actuation is the ESFAS protective function actuated by Tcold - High.

The Tcold - High Function 11.b would be required to be OPERABLE in MODES 1, 2, and 3 and in MODE 4 when the RCS is not being cooled by the Normal Residual Heat Removal System (RNS). This would ensure that the PRHR can be actuated in the event of a loss of the normal heat removal systems. In MODE 4 when the RCS is being cooled by the RNS, and in MODES 5 and 6, the PRHR is not required to provide the normal RCS heat sink.

The ESFAS initiates necessary safety systems, based upon the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the ESFAS, as well as specifying LCOs on other reactor system parameters and equipment performance.

TSs are required by 10 CFR 50.36 to include LSSS for variables that have significant safety functions. LSSS are defined by the regulation as where a limiting safety system setting is specified for a variable on which a safety limit has been placed; the setting must be chosen so that automatic protective actions will correct the abnormal situation before a safety limit is exceeded. The settings for automatic protection channels must be chosen to be more conservative than the Safety Analysis Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would occur.

Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires an LCO for actuation signals that form part of the primary success path and which functions or actuates to mitigate a design basis accident. The addition of the proposed new Tcold - High Function 11.b to TS Table 3.3.8-1 meets this criterion.

Therefore, the NRC staff finds the proposed change to add a new Tcold - High Function 11.b to TS Table 3.3.8-1 would continue tomeet 10 CFR 50.36 and is, therefore, acceptable.

3.5.6 Vendor Oversight Plan Summary The NRC staff evaluated whether the licensees oversight activities, as described in the Vogtle Electric Generating Plant, Units 3 and 4 Passive Residual Heat Removal (PRHR) Actuation Logic Modification Vendor Oversight Plan (VOP), SNC16282111VOP, Revision 3, Summary (hereafter referred to as the VOP Summary) meets the following criterion to Appendix B to 10 CFR Part 50:

Criterion III, Design Control

Criterion V, Instructions, Procedures, and Drawings

Criterion VII, Control of Purchased Material, Equipment, and Services

Criterion XI, Test Control

Criterion XVI, Corrective Action In addition, the NRC staff used the alternate review process (ARP) criteria in Revision 2 of Digital I&C (DI&C)-Interim Staff Guidance (ISG)-06, Digital Instrumentation and Control Licensing Process, Interim Staff Guidance (ML18269A259), to review the oversight activities described in the VOP Summary. Revision 2 of DI&C-ISG-06 defines the licensing process used to support the review of LARs associated with safety-related DI&C equipment modifications in operating plants. The ARP described in DI&C-ISG-06, Revision 2, allows the NRC staff to decide whether to approve an LAR after the system design is completed and evaluated, but before the system has been built and factory acceptance testing (FAT) completed where acceptability of the application-specific DI&C platform system is partially based on the licensees oversight and evaluation of the vendors DI&C system development process activities, as described in the licensees VOP Summary and VOP, Revision 3.

The VOP Summary describes each section of the Vogtle Electric Generating Plant, Units 3 and 4 Passive Residual Heat Removal (PRHR) Actuation Logic Modification Vendor Oversight Plan (VOP), SNC16282111VOP, Revision 3, and summarizes how the activities described in the VOP will ensure the licensees oversight of its vendor, Westinghouse Electric Corporation (WEC), during the development (e.g., software and design documentation) of the Vogtle, Units 3 and 4, PRHR Actuation Logic Change software modification. This includes design and testing deliverables that WEC will provide.

The licensees execution of the VOP, as described in the VOP Summary will verify that its vendor executes the project consistent with the VOP, and provides reasonable assurance that the as-developed and as-tested PRHR logic modifications will continue to meet the design and quality regulatory requirements of 10 CFR 50.55a(h), via IEEE Std 603-1991, and applicable criteria in Appendix B to 10 CFR Part 50. The NRC staff audited SNC1628211VOP, Vogtle Electric Generating Plant, Units 3 and 4 Passive Residual Heat Removal Actuation Logic Modification Vendor Oversight Plan, Revision 1, to identify details supporting the VOP Summarys description of vendor oversight activities and associated processes to perform these activities. The NRC staff documented its observations in the audit summary (ML25209A442).

The NRC staffs evaluation of the information within the VOP Summary to the applicable criteria in Appendix B to 10 CFR Part 50 is provided in Subsections 3.5.6.1 through 3.5.6.5 of this SE as described below.

3.5.6.1 Criterion III, Design Control Section 6 of the VOP Summary identifies performance measures and their acceptance criteria that will be included in the VOP. These performance measures are divided into the following three categories:

Critical Characteristics

Design Artifacts Programmatic Elements Critical Characteristics Section 6 of the VOP Summary describes critical characteristics as those important design, material, and performance characteristics of a system that, once verified, will provide reasonable assurance that the system will perform its intended critical functions. For the PRHR actuation logic change modification, the applicable critical characteristics are performance and cyber security. Examples of performance characteristics that will be validated include: (1) input ranges and setpoints, (2) outputs, output ranges, and data types, (3) features that are credited for surveillance testing or calibration, (4) response time requirements, and (5) central processing unit maximum load requirements. Cyber security critical characteristics will be addressed as part of the programmatic element.

For performance critical characteristics, the VOP Summary states that the licensee will verify:

(1) the logic functions being modified have been adequately included in the requirements traceability matrix (RTM), (2) modified logic functions perform as required, and (3) changes to the original design requirements and specifications for component modifications are identified as new requirements. In addition, the licensee will review segments of the RTM by examining the upstream and downstream document references for correct linkage.

Based on the description of oversight activities in the VOP Summary to verify the applicable critical characteristics for the PRHR actuation logic change modification, the NRC staff determined that the VOP Summary is adequate to meet the design control requirements of Criterion III to Appendix B of 10 CFR Part 50 and is, therefore, acceptable.

Design Artifacts Section 6 of the VOP Summary states that SNC will perform owners acceptance review of WEC design deliverables in accordance with B-GEN-ENG-038, Vogtle 3&4 Startup Engineering Change Procedure. This section describes generic design artifact oversight activities and oversight activities for specific documents. Section 6 of the VOP Summary states that documents and test reports developed by WEC and issued to the licensee will receive, as a minimum, an owners acceptance review. The licensee states that it is crediting the software development processes defined in the NRC approved WCAP-16096, Software Program Manual for the Common Q Systems, as modified by WCAP-15927, Design Process for AP1000 Common Q Safety Systems, under Westinghouses quality assurance (QA) program.

Section 6 of the VOP Summary identifies the following WEC documents that will be reviewed for owners acceptance:

Software Hazards Analysis

Failure Modes and Effects Analysis

Requirements Traceability Matrix

Software Test Plan Section 6 of the VOP Summary states that the licensee will perform owners acceptance of documents delivered by WEC to verify plant-specific action items identified in WEC platform topical reports are addressed in these documents.

Based on the description of the oversight activities for design artifacts in the VOP Summary, including performance of owners acceptance review for the WEC design deliverables of this project, the NRC staff determines that the VOP Summary is adequate to meet the design control requirements of Criterion III of Appendix B to 10 CFR Part 50 and is, therefore, acceptable.

VOP Change Process Section 3 of the VOP Summary states that VOP, Section 1.4, Revisions to the Vendor Oversight Plant, identifies the VOP as a Controlled Document. This section states that identified changes to the VOP require the following actions, prior to implementation:

Initiation of a condition report (CR) in the licensees corrective action program to track and document the approval, implementation, and communication of the change.

Review, approval, and administration of the change in accordance with SNC procedures B-GEN-ENG-038, NMP-ES-045-001, Technical Oversight Reviews of Engineering Products, and NMP-AD-025, Quality Assurance and Non-Quality Records Administration.

Review of the NRC safety evaluation approving the digital upgrade license amendment to confirm that the proposed VOP changes will not adversely impact the basis or requirements for NRC approval.

The NRC staff audited the VOP, Revision 3 and confirmed that the VOP change process included in the VOP, Section 1.4, is consistent with the VOP Summary. Based on the inclusion of the above conditions in the VOP and VOP Summary to govern changes to the VOP, the NRC staff determines that the licensees change control process for the VOP is adequate to meet the requirements of Criterion III of Appendix B to 10 CFR Part 50.

Based on the above, the NRC staff concludes that the oversight activities described in the VOP Summary are sufficient for the licensee to verify that there will be adequate design control during the PRHR actuation logic change modification to meet the requirements of Criterion III of Appendix B to 10 CFR Part 50 and is, therefore, acceptable. The NRC staffs evaluation of programmatic elements is in Section 3.5.6.3 of this SE.

3.5.6.2 Criterion V, Instructions, Procedures, and Drawing Section 3 of the VOP Summary states that the VOP is designed to be an umbrella document covering the range of activities in which [the licensee] is engaged to perform effective vendor oversight, and a hierarchy of [licensee] procedures ensure the effectiveness of vendor quality activities and products. Section 3 of the VOP Summary also states that the licensees Nuclear Development Quality Assurance Manual (NDQAM) implements Appendix B to 10 CFR Part 50 for Vogtle, Units 3 and 4. The NDQAM is implemented through the use of approved procedures, which provide written guidance for the control of quality related activities and provide for the development of documentation to provide objective evidence of compliance.

Section 3 of the VOP Summary identifies the primary implementing procedures for the NDQAM that are applicable to the PRHR actuation logic change modification. These procedures include:

NOS-201, Supplier Quality Program Evaluation

NOS-204, Supplier Audit/Survey Report Review

ND-QA-005, Quality Assurance Reviews ND-QA-008, Training and Qualification of Quality Assurance Personnel

ND-QA-019, Vogtle 3 &4 Supplier Qualification

NOS-202, Supplier Safety-Related program Audits

NOS-401, Supplier Quality Surveillance Section 3 of the VOP Summary identifies project management procedures that are used for the PRHR actuation logic change modifications, including:

NMP-GM-011, Procurement, Receipt, and Control of Materials and Services

NMP-ES-067, Major Project management

NMP-ES-067-004, Major Project Management Instruction

NMP-ES-067-005, Fast Track Project Instruction

NMP-ES-067-009, SNC Major Project Management Procurement Strategy Develop

NMP-ES-067-003, Graded Approach to Major Project Management

NMP-ES-067-006, Project Risk Management Section 3 of the VOP Summary identifies engineering and design control procedures that are used for the PRHR actuation logic change modifications, including:

NMP-ES-040-001, Preparation and Revision of Procurement Specifications for Engineered Components

NMP-ES-042, Design Input and Verification Process

NMP-ES-045, Design Authority

NMP-ES-045-001, Technical Oversight Reviews of Engineering Products

B-GEN-ENG-038, Vogtle 3&4 Startup Engineering Change Procedure

NMP-ES-095, Interface Procure for IP-ENG-001, Standard Design Process

NMP-GM-007, Acquisition and Development of Technology Solutions for Southern Nuclear

NMP-MA-014, Post Maintenance Testing/Post Modification Testing

NMP-MA-014-001, Post Maintenance Testing Guidance

NMP-GM-014, Cyber Security for Digital Plant Systems

B-GEN-CSEC-002, Cyber Security Team (CST) Implementation Instructions Section 3 of the VOP Summary identifies NMP-GM-002, Corrective Action Program, as the procedure that governs implementation of the licensees corrective action program.

Based on the identification and description of applicable implementing procedures for the (1)

NDQAM, (2) project management, (3) engineering and design control, and (4) corrective action program that will ensure the effectiveness of the vendor quality activities and products for the PRHR actuation logic change modification in the VOP Summary, the NRC staff finds that the licensee will implement vendor oversight activities for the PRHR actuation logic change modification in accordance with documented instructions and procedures that are consistent with the licensees QA program as described in the licensees NDQAM. Based on the above, the NRC staff concludes that the VOP Summary would continue to meet the requirements of Criterion V of Appendix B to 10 CFR Part 50 and is, therefore, acceptavno.

3.5.6.3 Criterion VII, Control of Purchase Materials, Equipment, and Services Section 2 of the VOP Summary identifies the vendor oversight activities that the licensee will perform to confirm that Westinghouse processes and products for the PRHR actuation logic change software modification meet contractual and regulatory obligations for the project. The licensee states in the VOP Summary that the results of the VOP will confirm that the activities related to software development coincide with the Westinghouse specified software lifecycle activities, as described in the Common Q Software Program Manual (SPM).

Section 6 of the VOP Summary identifies specific oversight activities that the licensee will perform to verify that (1) critical characteristics for the PRHR actuation logic change are met, (2) design artifacts produced by Westinghouse for the project are acceptable, and (3) programmatic elements of Westinghouses programs and processes for the project, as identified in the Common Q SPM, are adequately implemented.

Section 6 of the VOP Summary states that oversight of critical characteristics includes:

Conducting vendor audits and quality surveillances in accordance with the licensees NDQAM.

Conducting pre-planned on-site or remote vendor surveillances.

Conducting on-going quality and design-related review interactions and providing feedback to WEC.

Reviewing WEC design output documents.

Participating in FAT.

Observing or witnessing specific vendor testing activities.

Capturing issue in SNC/WEC corrective action programs.

Section 6 of the VOP Summary identifies examples of oversight activities and acceptance criteria for these design artifacts for each of the lifecycle phases within the scope of the VOP Summary and the VOP. These activities include:

Verification that the Vogtle, Units 3 and 4, specific requirements are correct, understandable, unambiguous, full the purchase specification, and are developed in accordance with the Common Q SPM, Section 10.2.1.

Verify requirements that are adopted without modification are validated using requirements phase independent verification and validation (IV&V) and design phase IV&V surveillance.

Verify requirements that are adopted without modification are validated by the FAT, including system validation test.

Verify (1) the requirements in the Software Requirements Specification are correct, understandable, unambiguous, fulfill the purchase specification, and are developed in accordance with the Common Q SPM, Section 10.2.2, and (2) each requirement is traceable to one or more system requirements.

Review the IV&V report on the software or logic design specifications.

Verify that the Software Hazard Analysis identifies the hazardous states and sequence actions that can cause the system to enter into a hazardous state.

Verify that WEC provides failure modes and effects analysis that demonstrate compliance with procurement specification.

Section 6 of the VOP Summary states that the licensee will conduct oversight activities and inspections for the WEC programs and processes relevant to the project as described in the Common Q SPM. These oversight activities include:

Review selected deliverables to validate the scope of documents revised to implement this change is consistent with the credited change process.

Review the FAT procedure.

Witness the FAT and the resolution of any anomalies.

Review the RTM for the project.

Review the Westinghouse Configuration Management Release Reports.

Sample software changes and verify that these changes are in compliance with the SPM for IV&V regression analysis.

Confirm that WEC documentation exists to show the IV&V tasks have been successfully performed, including documentation of (1) the software requirements traceability analysis, (2) any issues identified during the IV&V activities, and (3) resolution of any issues identified during the IV&V activities.

Section 9 of the VOP Summary states that the licensee will document vendor oversight activities to provide assurance that the licensee has been conducting oversight of Westinghouse through the system development lifecycle. The following methods will be used to document the licensees vendor oversight activities:

Formal audit plans/reports

Comments/feedback on design artifacts through the owner acceptance engineering process

Teleconference notes

Emails

Written correspondence between the licensee and Westinghouse Based on the above oversight activities, the NRC staff finds that the activities described in the VOP Summary adequately capture the oversight activities that will be performed to verify that Westinghouses processes and products for the PRHR actuation logic change software modification (1) are performed in accordance with the Common Q SPM, as modified by WCAP-15927, (2) meet procurement and regulatory requirements, and (3) PMS logic design changes specified in the LAR. The NRC staff also finds the surveillances of Westinghouse software and development processes, quality audits of Westinghouses activities in accordance with the Licensees NDQAM, and performance of acceptance reviews of Westinghouse design and IV&V artifacts, as described in the VOP Summary, will provide sufficient objective evidence of quality for the design outputs produced by Westinghouse for each phase of the PRHR actuation logic change software. Based on the above, the NRC staff concludes that the VOP Summary would contuniue to meet the requirements of Criterion VII of Appendix B to 10 CFR Part 50 and is, therefore, acceptable.

3.5.6.4 Criterion XI, Test Control Section 3 of the VOP Summary describes the licensees process and procedures that will be used to provide oversight of acceptance testing activities, including NMP-MA-014, Post Maintenance Testing/Post Modification Testing, and NMP-MA-014-001, Post Maintenance Testing Guidance. Section 6 of the VOP Summary identifies the following licensee oversight activities for factory acceptance testing and system validation activities:

Verify that validation and acceptance tests and reports required by the IV&V plan were successfully completed.

Review procedures for handling errors and anomalies encountered during the IV&V reviews and tests.

Verify that regression analysis is performed in accordance with the Common Q SPM.

Review and verify the Software Test Pan and the system validation are in compliance with the design requirements.

Review the FAT procedure.

Witness the FAT and the resolution of any anomalies.

The NRC staff finds that the above description of the licensees oversight activities are adequate to verify that Westinghouse (1) performed regression analysis to identify the extent of re-testing needed, (2) conducted all testing required to demonstrate the PRHR actuation logic change modification meets requirements and acceptance limit contained in applicable design documents, and (3) test results are adequately documented. Based on the above, the NRC staff concludes that the VOP Summary meets the requirements of Criterion XI of Appendix B to 10 CFR Part 50 and is, therefore, acceptable.

3.5.6.5 Criterion XVI, Corrective Actions Section 8 of the VOP Summary, Perform Corrective Actions, states that Condition Reports (CRs) will be the entry point into the corrective action program to document vendor performance or quality that is in question. The following conditions, as a minimum, trigger the initiation of a CR:

Westinghouse noncompliance with the Westinghouse quality program or software processes

Nuclear safety may be adversely impacted if the digital software or information item is installed and operated

Unit generation may be adversely impacted if the digital software or information item is installed and operated

Digital software or information item quality cannot be assured

Digital software or information item quality cannot be assured without a significant project delay Section 8 of the VOP Summary states that if the licensee identifies performance issues, oversight would be enhanced to include:

Periodic meetings to discuss and resolve issues

Additional technical reviews or surveillances

Management intervention

Stop work and implement recovery plan Based on the above, the NRC staff determines that the minimum conditions that would trigger initiation of a CR identified in the VOP Summary are adequate to ensure the potential conditions adverse to quality will be identified and corrected. The NRC staff also determined the description of measures that would be taken to enhance the oversight of WEC should performance issues arise will support resolution of performance issues, minimize risks associated with these performance deficiencies, and reduce the likelihood that conditions adverse to quality will occur. Based on the above, the NRC staff concludes that the VOP Summary meets the requirements of Criterion XVI of Appendix B to 10 CFR Part 50, and is therefore, acceptable.

3.6 Technical Conclusion The NRC staff finds the proposed editorial changes to the TSs acceptable, because the proposed editorial changes do not alter the intent, scope, or requirements in the TS, but the editorial changes serve to improve its presentation and usability.

The NRC staff finds the changes to TS 3.3.8 acceptable, because they meet 10 CFR 50.36(c)(2)(C), which is Criterion 3 for LCO inclusion in TS. The condition Low-2 SG NR water level coincident with Low-2 startup feedwater flow coincident with High cold leg temperature (Tcold - High) results from the coincidence of two of the four divisions of SG NR water level below the Low-2 setpoint, after a preset time delay, coincident with a Low-2 startup feedwater flow for both SGs coincident with High cold leg temperature (Tcold - High) in either loop. The NRC staff finds that the changes to TS 3.3.8 continue to meet GDCs 13, 20, and 34; therefore, the NRC staff finds the changes to TS 3.3.8 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 July 14, 2025. On July 21, 2025, the State official confirmed that the State of Georgia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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, and there has been no public comment on such finding published in the Federal Register on May 14, 2025 (90 FR 20519). 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

N. Soliz, NRR N. Carte, NRR K. West, NRR Principal Contributors:

J. Ambrosini, NRR D. Zhang, NRR A. Stubbs, NRR Dated: August 8, 2025.

ML25199A048 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/SCPB/BC NRR/DEX/EICB/BC NAME JLamb Zeleznock (ABaxter for)

MValentin SDarbali for FSacko DATE 07/21/2025 08/04/2025 07/29/2025 07/28/2025 OFFICE NRR/DSS/STSB/BC NRR/DSS/SNSB/BC NRR/DRO/IQVB/BC NRR/DORL/LPL2-1/BC NAME SMehta NDifrancesco KKavanagh MMarkley DATE 08/01/2025 07/29/2025 07/29/2025 08/08/2025 OFFICE NRR/DORL/LPL2-1/PM NAME JLamb DATE 08/08/2025