ML22116A084

Issuance of Amendment Nos. 215 and 198, Regarding Revision to Technical Specification 3.7.2, Main Steam Isolation Valves (Msivs)

ML22116A084
Person / Time
Site:	Vogtle
Issue date:	07/05/2022
From:	John Lamb Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Lamb, J.
References
EPID L-2021-LLA-0178
Download: ML22116A084 (43)

Text

July 5, 2022 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 215 AND 198, REGARDING REVISION TO TECHNICAL SPECIFICATION 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVS)

(EPID L-2021-LLA-0178)

Dear Ms. Gayheart:

The Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 215 to Renewed Facility Operating License NPF-68 and Amendment No. 198 to Renewed Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the license and technical specifications (TSs) for each unit in response to your application dated September 30, 2021, as supplemented by letters dated January 13 and February 25, 2022.

The amendments revise TS 3.7.2, Main Steam Isolation Valves (MSIVs), Limiting Condition of Operation (LCO), to require four MSIVs and associated actuators and bypass valves be Operable in MODE 1, and in MODES 2 and 3, with exceptions. The amendments to TS 3.7.2 also add and modify the Conditions and Required Actions, update the existing Surveillance Requirement (SR), and add a new SR to reflect the change in the LCO requirements.

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

Sincerely,

/RA/

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

Enclosures:

1. Amendment No. 215 to NPF-68

2. Amendment No. 198 to NPF-81

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

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 215 Renewed License No. NPF-68 1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 30, 2021, as supplemented on January 13 and February 25, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

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

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented prior to entering MODE 3 during the start up from the spring 2023 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

Attachment:

Changes to License No. NPF-68 and the Technical Specifications Date of Issuance: July 5, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.07.05 15:43:47 -04'00'

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 198 Renewed License No. NPF-81

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 30, 2021, as supplemented January 13 and February 25, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

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

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 198, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented prior to entering MODE 3 during start up from the fall 2023 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

Attachment:

Changes to License No. NPF-81 and the Technical Specifications Date of Issuance: July 5, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.07.05 15:44:31 -04'00'

ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 TO LICENSE AMENDMENT NO. 215 RENEWED FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 198 RENEWED FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of the Licenses and the Appendix A, Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. NPF-68, page 4 License No. NPF-68, page 4 License No. NPF-81, page 3 License No. NPF-81, page 3 TSs TSs 3.7.2-1 3.7.2-1 3.7.2-2 3.7.2-2 3.7.2-3 Renewed Operating License NPF-68 Amendment No. 215 (1)

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

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

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

(4)

Deleted (5)

Deleted (6)

Deleted (7)

Deleted (8)

Deleted (9)

Deleted (10)

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

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training and response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for Implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy Renewed Operating License NPF-81 Amendment No. 198 (2)

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this license; (3)

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

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

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

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

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

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

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 198 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

MSIVs 3.7.2 Vogtle Units 1 and 2 3.7.2-1 Amendment No. 215 (Unit 1)

Amendment No. 198 (Unit 2) 3.7 PLANT SYSTEMS 3.7.2 Main Steam Isolation Valves (MSIVs)

LCO 3.7.2 Four MSIVs and their associated actuator trains and associated bypass valves shall be OPERABLE.

APPLICABILITY:

MODE 1, MODES 2 and 3 except when one MSIV and one bypass valve in each steam line are closed.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One MSIV with one actuator train inoperable A.1 Restore MSIV actuator train to OPERABLE status.

7 days B.

Two MSIV actuator trains inoperable on different MSIVs and on different trains.

B.1 Restore one MSIV actuator train to OPERABLE status.

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

Two MSIV actuator trains inoperable on different MSIVs on the same train.

C.1 Restore one MSIV actuator train to OPERABLE status.

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

Two MSIV actuator trains inoperable on the same MSIV.

D.1 Declare the affected MSIV inoperable.

Immediately E.

Three or more MSIV actuator trains inoperable.

OR Required Action and associated Completion Time of Condition A, B, or C not met.

E.1 Declare each affected MSIV inoperable.

Immediately (continued)

MSIVs 3.7.2 Vogtle Units 1 and 2 3.7.2-2 Amendment No. 215 (Unit 1)

Amendment No. 198 (Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F.

One MSIV inoperable in MODE 1.

F.1 Restore MSIV to OPERABLE status.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> G. Required Action and associated Completion Time of Condition F not met.

G.1 Be in MODE 2.

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

---

NOTES----------

Separate Condition entry is allowed for each bypass valve.

One or more bypass valves inoperable.

H.1 Close or isolate the bypass valve.

AND H.2 Verify the bypass valve is closed or isolated.

7 days Once per 7 days thereafter I.

---

NOTES----------

Separate Condition entry is allowed for each MSIV.

One or more MSIVs inoperable in MODE 2 or

3.

I.1 Close or isolate MSIV.

AND I.2 Verify MSIV is closed or isolated.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AND Once per 7 days thereafter J.

Required Action and associated Completion Time of Condition H or I not met.

J.1 Be in MODE 3.

AND J.2 Be in MODE 4.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours

MSIVs 3.7.2 Vogtle Units 1 and 2 3.7.2-3 Amendment No. 215 (Unit 1)

Amendment No. 198 (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1

---

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

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV and bypass valve is within limits on an actual or simulated actuation signal.

In accordance with the INSERVICE TESTING PROGRAM SR 3.7.2.2

---

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

Only required to be performed in MODES 1 and 2.

Verify each actuator train actuates the MSIV to the isolation position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 198 TO RENEWED FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

By application dated September 30, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21274A073), as supplemented on January 13 and February 25, 2022 (ADAMS Accession Nos. ML22014A417 and ML22056A433, respectively),

Southern Nuclear Operating Company, Inc. (SNC, the licensee), requested revisions to the technical specifications (TSs) for the Vogtle Electric Generating Plant (Vogtle), Units 1 and 2.

The proposed amendments revise TS 3.7.2, Main Steam Isolation Valves (MSIVs), Limiting Condition of Operation (LCO), to require four MSIVs and associated actuators and bypass valves be Operable in MODE 1, and in MODES 2 and 3 with exceptions. The amendments to TS 3.7.2 also add to and modify the Conditions and Required Actions (RAs), update the existing Surveillance Requirement (SR), and add a new SR to reflect the change in the LCO requirements.

The supplements dated January 13 and February 25, 2022, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published the Federal Register on December 28, 2021 (86 FR 73819).

2.0 REGULATORY EVALUATION

2.1 Background

Vogtle, Units 1 and 2, have two main steam isolation valves and two main steam isolation bypass valves per main steam line (MSL) rather than the typical single MSIV and bypass valve per steam line. The valves are situated as close to the containment structure as practical but

outside the containment building. The redundancy of the MSIVs and bypass isolation valves in each line provides positive shutoff with minimum leakage during possible line-break situations either upstream or downstream of the valves. Whereas many other pressurized-water reactor facilities must rely on non-safety-grade turbine stop valves to mitigate the consequences of a pipe break when considering a single active failure, Vogtle, Units 1 and 2, because of its redundant MSIVs, do not rely on the stop valves even with a single failure.

2.2

System Description

The SNC letter dated September 30, 2021, stated the following.

The main steam supply system includes the following major components:

A. Main steam piping from the steam generator outlet steam nozzles to the main turbine stop valves.

B. Two MSIVs and two MSIV bypass valves per main steam line.

C. Main steam safety valves.

D. Power-operated atmospheric relief valves.

The power operated atmospheric relief, safety, MSIVs, and MSIV bypass valves are located outside the containment and are installed as close as possible to the containment wall.

Each main steam line contains two MSIVs in parallel with two MSIV bypass valves. These valves isolate the secondary side of the steam generators to deal with leakage and malfunction and to prevent the uncontrolled blowdown of two steam generators and isolate non-safety related portions of the main steam system. This constitutes the design function of the main steam isolation system.

Turbine bypass valves are provided between the MSIVs and turbine-generator stop valves.

Steam from each of four steam generators enters the high-pressure turbine through four stop valves and governing control valves. Crossties are provided upstream of the turbine stop valves to provide pressure equalization with one or more stop valves closed. Each stop valve is controlled by an electrohydraulic actuator, so that the stop valve is either fully open or fully closed. The function of the stop valves is to shut off the steam flow to the turbine, when required.

2.3 Regulatory Requirements Section 182a. of the Atomic Energy Act of 1954, as amended (Act), requires that applications for nuclear power plant operating licenses include TSs and that the TSs be part of the license.

These TSs are derived from the plant safety analyses.

Section 50.36 to Title 10 to the Code of Federal Regulations (10 CFR) contains the requirements for the content of the TS. Pursuant to 10 CFR 50.36, TSs for reactors are required to include items in the following five specific categories related to station operation: (1) safety limits and limiting safety system settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls.

Section 50.36(a)(1) to 10 CFR requires each applicant for a license to include a summary statement of the bases or reasons for proposed TS; however, the bases shall not become part of the TSs.

Section 50.36(c)(2) to 10 CFR states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility and that when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met.

Section 50.36(c)(3) to 10 CFR states that SRs 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 LCOs will be met.

Section 50.49 to 10 CFR, Environmental qualification of electric equipment important to safety for nuclear power plants, requires, in part, licensees to establish a program for qualifying the electric equipment important to safety. The electric equipment under the scope of this section includes safety-related equipment, non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified by the safety-related equipment, and certain post-accident monitoring equipment. For compliance with 10 CFR 50.49, the safety-related equipment is that equipment that is relied upon to remain functional during and following design-basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in 10 CFR 50.34(a)(1), 50.67(b)(2), or 100.11, as applicable. The regulation 10 CFR 50.49(e)(1) requires that the time-dependent temperature and pressure at the location of the electric equipment important to safety must be established for the most severe design basis accident during or following which this equipment is required to remain functional. The regulation 10 CFR 50.49(e)(2) requires that humidity during design-basis accidents must be considered. The regulation 10 CFR 50.49(e)(3) requires that the composition of chemicals used must be at least as severe as that resulting from the most limiting mode of plant operation (e.g.,

containment spray, emergency core cooling, or recirculation from containment sump). If the composition of the chemical spray can be affected by equipment malfunctions, the most severe chemical spray environment that results from a single failure in the spray system must be assumed. The regulation 10 CFR 50.49(e)(4) requires that the radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects. The regulation 10 CFR 50.49(c) states that requirements for (1) dynamic and seismic qualification of electric equipment important to safety, (2) protection of electric equipment important to safety against other natural phenomena and external events, and (3) environmental qualification of electric equipment important to safety located in a mild environment are not included within the scope of 10 CFR 50.49.

The regulation 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, states, in part, that valves that are within the scope of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) must meet the inservice test (IST) requirements set forth in the ASME OM Code (except design and access provisions), and that IST requirements for valves that are within the

scope of the ASME OM Code but are not classified as ASME Boiler and Pressure Vessel Code (BPV Code) Class 1, Class 2, or Class 3 may be satisfied as an augmented IST program.

The regulation 10 CFR 50.55a(h), Protection and safety systems, requires, as applicable to Vogtle, Units 1 and 2, that nuclear power plants (NPPs) meet the criteria of Institute of Electrical and Electronic Engineers (IEEE) Standard (STD) 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, which establishes the minimum requirements for the safety-related functional performance of protection systems for NPPs, including the general functional and single failure criteria.

The regulation 10 CFR 50.63, Loss of all alternating current power, states, in part, that the plant must be able to withstand for a specified duration and recover from a station blackout (SBO).

Part of the regulatory framework for requesting this licensing action is based on 10 CFR Section 100.11, Determination of exclusion area, low population zone, and population center distance, which requires, in part, that the licensee determine:

(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem [roentgen equivalent man] or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

Appendix A, to 10 CFR Part 50, General Design Criteria [GDC] for Nuclear Power Plants, requires, in part, that a proposed facility must include principal design criteria that establish minimum requirements. The GDC applicable to TS 3.7.2 include:

GDC 2, Design bases for protection against natural phenomena, states, in part, that structures, systems, and components (SSCs) important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.

GDC 4, Environmental and dynamic effects design bases, states, in part, that SSCs important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents [LOCAs].

GDC 5, Sharing of structures, systems, and components, states that the SSCs important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

GDC 16, Containment design, states, in part, that the reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 17, Electric power systems, requires, in part, that an onsite and offsite electric power system shall be provided to permit functioning of SSCs important to safety. The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

GDC 34, Residual heat removal, states, in part, that a system to remove residual heat shall be provided, and 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.

2.4 Regulatory Guidance In determining the acceptability of revising TS 3.7.2, the NRC staff used plant-specific licensing basis information as well as the accumulation of generically approved guidance in the following documents.

Regulatory Guide (RG) 1.195, Revision 0, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors (ADAMS Accession No. ML031490640).

RG 1.29, Seismic Design Classification, Revision 3 (ADAMS Accession No. ML003739983).

RG 1.117, Tornado Design Classification, Revision 1 (ADAMS Accession No. ML003739346).

RG 1.155, Station Blackout, Revision 0 (ADAMS Accession No. ML003740034).

NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 4, dated April 2012 (ADAMS Accession No. ML12100A222).

The following sections of NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, were used to evaluate this license amendment request (LAR).

o SRP Chapter 15, Section 15.1.5, Appendix A, Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR [Pressurized-Water Reactor], Revision 2 (ADAMS Accession No. ML052350118), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR Part 100.

o SRP Chapter 15, Section 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR), Revision 2 (ADAMS Accession

No. ML052350149), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR Part 100.

o SRP Chapter 6, Section 6.2.1.4, Mass and Energy Release Analysis for Postulated, Secondary System Pipe Ruptures, Revision 2 (ADAMS Accession No. ML070620010), provides guidance to evaluate sources of energy, mass and energy release rate, and single-failure analyses performed for steam-and feedwater-line ruptures for containment functional design.

o SRP Chapter 10, Section 10.3, Main Steam Supply System, Revision 4 (ADAMS Accession No. ML070380206), provides guidance to evaluate the main steam supply system from the containment up to the turbine stop valve.

The single-failure criterion for nuclear power plants is the requirement for safety-related structures, systems, and components (SSCs) used to mitigate abnormal operational occurrences and design basis accidents (DBAs) to have sufficient redundancy in components and features, among other things, such that the safety function(s) for any such SSC can be accomplished assuming any single failure. Single failure is defined in Appendix A, of 10 CFR Part 50, and the single-failure criterion is stated in GDCs 17, 34, 35, 38, 41, and 44. Plants are designed and licensed to meet the single-failure criterion. Plants are normally operated with the requirement that the single-failure criterion is being met in that the TSs contain LCOs that require all necessary SSCs that meet one or more of the four criteria in 10 CFR 50.36(c)(2)(ii) to be operable.

Plant TSs are formulated to preserve the single-failure criterion for SSCs described in the updated final safety analysis report that are relied upon in the DBA analyses. The single-failure criterion is preserved by specifying LCOs that require all redundant components of safety-related systems to be operable. When the required redundancy is not maintained, either due to equipment failure or a maintenance outage, the TSs require plant shutdown or a remedial action to be taken within a Completion Time (CT), which is a temporary relaxation of the single-failure criterion. The specified CT provides a limited time, consistent with overall system reliability and risk considerations, to fix the equipment or otherwise make it operable.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes

The proposed amendment revises TS 3.7.2. The current TS LCO requires two MSIV systems per main steam line be Operable in MODE 1, and in MODES 2 and 3 with exceptions. The proposed amendment changes TS 3.7.2, LCO, to require four MSIVs and associated actuators and bypass valves be Operable in MODE 1, and in MODES 2 and 3 with exceptions. The proposed amendment changes the Conditions and RAs to incorporate the change in the LCO scope. The proposed amendment updates the existing SR and adds a new SR to reflect the change in the LCO requirements.

In the current TS at Vogtle, Units 1 and 2, an MSIV system is defined as an MSIV and its related bypass valve. An MSIV system is either a Train A or Train B system, that is, the signals to the MSIV and its related bypass valve are delivered from the same train (steam line isolation (SLI)-A or SLI-B).

In its letter dated September 30, 2022, SNC stated:

SNC proposes to change the licensing basis to eliminate the requirement for the automatic closure function from the outboard MSIV in each main steam line. The change to the plant design will be implemented upon approval of this licensing change. Both MSIVs will remain in the main steam line, in their current location.

The inboard MSIV will have a new dual train actuator installed with redundant actuation signals and redundant power supplies. The existing actuator will remain on the outboard valve (former MSIV) and will be capable of manual valve closure. Automatic closure of the outboard valve will be precluded by removal of the automatic actuation signal and depressurization of the nitrogen accumulator.

The outboard valve will still be capable of manual closure using the existing actuator upon charging of the nitrogen accumulator. The MSIV bypass valves remain in place with their existing closure signals. The non-safety related valves downstream of the outboard valve are not altered by this change.

3.1.1 Proposed LCO 3.7.2 The proposed changes to LCO 3.7.2 are marked below.

SNC proposed to add dual train actuators and the bypass valves to the LCO, Conditions and SRs.

3.1.2 Proposed Deletion of NOTE The proposed deletion of a NOTE for the ACTIONS for TS 3.7.2 is marked below.

3.1.3 Proposed Conditions A through E The licensee proposed new Conditions A through E as marked below.

CONDITION REQUIRED ACTION COMPLETION TIME A. One MSIV with one actuator train inoperable A.1 Restore MSIV actuator train to OPERABLE status.

7 days

B. Two MSIV actuator trains inoperable on different MSIVs and on different trains.

B.1 Restore one MSIV actuator train to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C. Two MSIV actuator trains inoperable on different MSIVs on the same train.

C.1 Restore one MSIV actuator train to OPERABLE status.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. Two MSIV actuator trains inoperable on the same MSIV.

D.1 Declare the affected MSIV inoperable.

Immediately E. Three or more MSIV actuator trains inoperable.

OR Required Action and associated Completion Time of Condition A, B, or C not met.

E.1 Declare each affected MSIV inoperable.

Immediately 3.13.4 Current Condition A - Proposed Condition F The licensee proposed to change the current Condition A to reflect the change from two MSIV systems per steam line to one MSIV per steam line and to re-label Condition A as Condition F.

In addition, the license proposed to delete the allowance for more than one MSL to have an inoperable MSIV system. SNC proposed the CT for restoration of the inoperable MSIV to be changed from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee proposed to eliminate the risk-informed completion time (RICT).

3.1.5 Current Condition B The Condition B is proposed to be deleted as marked below.

3.1.6 Current Condition C - Proposed Condition G The current Condition C is proposed to be re-labeled to Condition G and now applies only to Condition F as marked below.

3.1.7 Current Condition D The current Condition D is proposed to be deleted as marked below.

3.1.8 Proposed Condition H SNC proposed a new Condition H to be added to address an inoperable bypass valve as marked below. A separate Condition entry would be allowed for each bypass valve because it operates independently of the other bypass valves.

3.1.9 Current Condition E - Proposed Condition I SNC proposed to re-label current Condition E to Condition I as marked below. The licensee proposed to add a NOTE for separate Condition entry to Condition I. The Condition is changed to reflect the change from two MSIV systems per steam line to one MSIV per steam line. The licensee proposed RA I.1 be added to close or isolate the inoperable MSIV. The MSIV can be isolated by closing the outboard valve in the same MSL. This ensures the MSL flow is isolated supporting the assumptions in the safety analysis. SNC proposed to change the CT from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee proposed Condition I.2 be modified to reflect the change from two MSIV systems per steam line to one MSIV per steam line. It requires verification that the MSIV is closed or isolated.

3.1.10 Current Condition F - Proposed Condition J The licensee proposed to re-label current Condition F to Condition J. As marked below, SNC proposed that the reference to the previous Conditions D and E be changed to Conditions H and I.

JF. Required Action and associated Completion Time of Condition HD or IE not met.

JF.1 Be in MODE 3.

AND JF.2 Be in MODE 4.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours

3.1.11 Proposed Changes to SR 3.7.2.1 SNC proposed to change SR 3.7.2.1 to reflect the change from an MSIV system to an MSIV and bypass valve.

3.1.12 Proposed SR 3.7.2.2 SNC proposed to add SR 3.7.2.2 to address the dual train actuators on the MSIVs as marked below. The SR is needed to provide assurance that the actuators are Operable per the LCO.

3.1.13 Proposed TS Bases The licensee provided proposed TS bases. The NRC staff used the TS bases for informational purposes only. The NRC staff did not approve the TS Bases.

3.2 Change to MSIVs In its letter dated February 25, 2022, the licensee stated that the function of the MSIVs is to close in the event of a steam line break to prevent an uncontrolled blowdown of the steam generators (SGs). SNC said that based on the successful performance of TS SR 3.7.2.1 for the last 15 plus years, the MSIVs at Vogtle, Units 1 and 2, are very reliable in performing their specified safety function. SNC stated there have been six inadvertent trips since 2012 related to the inadvertent closure of one MSIV in a steam line. SNC stated that the six MSIV failures were due in part to the large number of single-point vulnerability (SPV) subcomponents associated with each valve.

The licensee estimated that there are approximately 40 subcomponents associated with the current actuator design which if failed would result in the unplanned closure of an MSIV. These subcomponents include dump solenoids, hydraulic seals, electrical connectors, relays, fuses, microswitches, and pressure retaining O-rings. The licensee stated that this reduction to one MSIV will improve plant reliability by preventing plant trips and/or transients associated with these SPVs.

3.3 IST Program The licensees IST Program Update dated October 23, 2017 (ADAMS Accession No. ML17298A197), indicates that the Code of Record for the Vogtle, Units 1 and 2, Fourth 10-Year IST Program interval is the 2004 Edition through 2006 Addenda of the ASME OM Code as incorporated by reference in 10 CFR 50.55a. In the LAR dated September 30, 2021, SNC stated that the IST Program and Inservice Inspection Program will be evaluated in accordance with 10 CFR 50.55a to determine what, if any, program changes are required with the programs updated accordingly. In its letter dated February 25, 2022, the licensee stated that as required by 10 CFR 50.55a(f) and the ASME OM Code, Subsection ISTA, General Requirements, paragraph ISTA-1100, Scope, and Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-1100, Applicability, it will continue to maintain all MSIVs in the IST Program following implementation of the LAR.

Further, SNC states that the MSIVs and MSIV bypass valves are currently included in the IST Program and will remain in the IST Program as identified below.

MSIV Valves MSIV Bypass Valves 1/2-HV-3006A&B 1/2-HV-13005A&B 1/2-HV-3016A&B 1/2-HV-13006A&B 1/2-HV-3026A&B 1/2-HV-13007A&B 1/2-HV-3036A&B 1/2-HV-13008A&B 3.4 Modification of the Main Steam Piping System and Its Supports Including Snubbers In its letter dated February 25, 2022, the licensee stated that no snubbers were added or removed for the implementation of this LAR. Also, the snubber loadings on existing snubbers are not impacted due to this LAR. The LAR implementation will have no impact on snubber hot and cold settings or the Snubber Inservice Examination and Testing Program because the effect of the new actuators on the pipe support loads is localized, and there are no snubbers on the sections of piping being modified. Further, the licensee stated that after implementation of this LAR, pipe stresses will remain within the applicable ASME BPV Code,Section III, Class 2 design requirements.

In its letter dated February 25, 2022, SNC stated that changes due to implementation of this LAR will have no impact on the extent of no-break zone piping identified in Section 3.6 of the Vogtle, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR). The boundaries of the no-break zone piping remain as described in the UFSAR. The licensee stated that design margins for the main steam line piping continue to meet the Vogtle Units 1 and 2, UFSAR requirements after implementation of the actuator replacement. Further, SNC stated that the configuration changes do not impact accessibility for inspections of the main steam line piping, including valves and/or supports, etc. Thus, the design criteria and the augmented inservice examination requirements set forth in the Vogtle, Units 1 and 2, UFSAR Section 3.6.1 and Table 3.6.1-3 are unaffected by this change.

Based on the information provided by the licensee in the submittal dated September 30, 2021, and the supplements, the NRC staff finds that the changes resulting from this LAR will not adversely impact the performance of these applicable valves or their status in the Vogtle, Units 1 and 2, IST Program. The NRC staff also finds that SNC will continue to meet 10 CFR 50.55a(f)(4) after implementation of this LAR with respect to these applicable valves. In addition, the NRC staff finds that the licensee has adequately demonstrated that the LAR

implementation will not adversely impact the snubber hot and cold settings nor the Snubber Inservice Examination and Testing Program. The licensee clarified that the new and modified main steam line will remain within the applicable ASME BPV Code,Section III, Class 2 stress design requirements and the proposed configuration changes will not adversely impact the extent of the no-break zone piping. Further, the Vogtle, Units 1 and 2, UFSAR design criteria and the augmented inservice examination requirements are not affected by the changes resulting from this LAR.

3.5 Evaluation of Risk Insights By letter dated August 8, 2017 (ADAMS Accession No. ML15127A669), the NRC issued Amendment Nos. 188 and 171 to Vogtle, Units 1 and 2, that modified the TSs to permit use of Risk-Informed Completion Times (RICTs) in accordance with Topical Report Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specification Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A (ADAMS Accession No. ML12286A322). Amendment Nos. 188 and 171 replaced the associated CT with a reference to a licensee-controlled document for selected TS action statements, and the required CTs in the licensee-controlled documents are managed in accordance with SNCs RICT Program.

According to the SNC letter dated September 30, 2021, SNC performed a probabilistic assessment to provide risk insights for the proposed TS changes and new CTs for TS 3.7.2, Conditions A, B, C, and F. However, SNC stated that the LAR was not a risk-informed request.

In the SNC letter dated September 30, 2021, SNC stated:

The probabilistic analysis examined two sets of risk metrics: the expected change in annual average CDF [core damage frequency] and LERF [large early release frequency], and the Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Release Probability (ICLERP).

The assessment determined that the proposed design modification results in a small increase in CDF and LERF, and the expected impact was found to be significantly below a 1E-06/yr [year] change in CDF and 1E-07/yr change in LERF. This criterion represents a very small increase in risk per Regulatory Guide 1.174. In addition, the assessment did not result in significant risk insights.

The NRC staff considered the risk insights provided by SNC. However, because the LAR was not a risk-informed request, the NRC staff did not rely on the risk-informed insights in its regulatory decision. The NRC staff notes, however, that the risk insights do not adversely impact its conclusions concerning the acceptability of the engineering and technical basis.

3.6 Radiological Consequences The NRC staff has reviewed current licensing basis design basis accidents (DBAs) as they are affected by the change from two TS-required MSIVs per steam line to one TS-required MSIV per steam line. The licensee stated, that with respect to radiological dose consequence analysis, the current analysis of record (AOR) DBAs which could be potentially affected by the changes requested in the LAR are the Steam Generator Tube Rupture (SGTR) and the Main

Steam Line Break (MSLB) DBAs. The SGTR and MSLB DBAs remain bounded in their current form with implementation of the changes requested in the LAR.

3.6.1 Steam Generator Tube Rupture The evaluation of the radiological consequences of a postulated SGTR assumes that the faulted steam generator (SG) encounters a double-ended break of one SG tube. The accident analysis assumptions include reactor operation at full power with a limited number of defective fuel rods.

In the analysis, there is a coincident loss of offsite power (LOOP), and the release to the environment is via the SG power-operated atmospheric relief valve on the faulted SG. In the accident sequence, the release via the atmospheric relief valve is terminated at 16 minutes when the associated block valve is manually closed.

The implementation of this LAR, with operation of a single MSIV per steam line, opens the possibility of the path for release due to a single failure becoming a release via the MSL with the failure to close of the MSIV associated with the failed SG. Immediate actions associated with SGTR require isolation of the affected SG, including closing of the associated MSIV.

In the current AOR, the release rate from the ruptured SG is provided in the Updated Final Safety Analysis Report (UFSAR) Figure 16.6.3-10 Steam Generator Tube Rupture Ruptured SG Atmospheric Mass Releases. In this figure, the peak release rate from the atmospheric relief valve is approximately 240 pounds mass per second (lbm/sec). In the licensees response dated February 25, 2022, to a request for additional information (RAI), the licensee states that a release rate of 255 lbm/sec (918,900 pounds mass per hour (lbm/hr)) is used in the calculation for the SGTR based on the most limiting single failure with respect the dose analysis is a failed open atmospheric relief valve (ARV) on the ruptured SG. The licensee stated that the failed open ARV bounds the ruptured SG MSIV failure to close scenario.

In response to a Nuclear Safety Advisory Letter, NASL 06-15, issued by Westinghouse, SNC performed an audit of the main steam system to identify the location and magnitude of branch line steam flows downstream of the MSIVs. SNC evaluated the components downstream of the MSIVs which would ultimately be the pathways for a release to the environment with a failed open MSIV. The seven identified potential pathways for release have a combined flow rate of 164 lbm/sec. Also, in its letter dated February 25, 2022, the licensee stated that in the event of the inboard MSIV failing to close remotely, the outboard valve will be locally manually closed within 16 minutes after the attempt to isolate the ruptured SG MSIV.

For the current AOR to remain bounding, the release rate and total mass released from a failed open ARV would need to exceed the total of the summed release rates and total mass released from the pathways downstream of a failed MSIV. Information provided in the current UFSAR, the LAR, and in the subsequent response to an RAI demonstrates that the current AOR, which models the release path during a SGTR to be via a failed open power operated atmospheric relief valve, has a release rate of 240 lbm/s, and will be isolated 16 minutes after the valve fails to close. The leak rate via the failed open MSIV to the branch steam lines totals 164 lbm/s and will be isolated via manual closure of the second MSIV 16 minutes after the attempt to isolate the rupture SG MSIV. The release rate and total integrated release via the ARV exceeds that of calculated release via a failed open MSIV

The current AOR remains limiting with respect to release rates and total integrated release to the environment. Therefore, the NRC concludes that licensees radiological consequence analysis is sufficient for SG tube rupture based on the analysis a MSIV failing to close being within the bounding analysis of a failed open ARV.

3.6.2 Main Steam Line Break The evaluation of the radiological consequences of a postulated main steam line break (MSLB) outside containment assumes that the reactor has been operating with a TS limit of defective fuel and leaking SG tubes for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant. Following the rupture, the auxiliary feedwater to the faulted loop is isolated and the SG is allowed to steam dry.

Radionuclides present in the primary coolant are released to the faulted SG via leaking tubes and are assumed to be released directly to the environment.

Section 15.1.5.3.1.3 UFSAR Identification of Leakage Pathways and Resultant Leakage Activity states:

For evaluating the radiological consequences due to a postulated MSLB, the activity released from the affected steam generator (steam generator connected to the broken steam line) is released directly to the environment. The unaffected steam generators are assumed to continually discharge steam and entrained activity via the safety and relief valves up to the time initiation of the RHR system can be accomplished.

All activity is released to the environment with no consideration given to radioactive decay or cloud depletion by ground deposition during transport to the exclusion area boundary and low population zone. The resultant radiological consequences of the current AOR represent the most conservative estimate of the potential integrated dose due to the postulated MSLB.

In the AOR for steam system piping failure, the faulted SG is not isolated and blows down completely with no credit for MSIV closure. Implementation of the changes described in this LAR would have no impact on the current DBA associated with a MSLB. Therefore, the current AOR remains bounding with respect to the radiological consequence of this DBA.

3.6.3 Radiological Consequences Conclusion The NRC staff reviewed the current licensing basis and the assumptions used by SNC to assess the changes requested in the LAR. Based on the above, the NRC staff finds these changes do not affect any previously approved DBA radiological accident dose analyses and are, therefore, acceptable.

3.7 Environmental Qualification (EQ) of Electrical Equipment Important to Safety The NRC staff reviewed the submittal to determine the impact of the proposed change on the environmental qualification (EQ) of electrical equipment. In the LAR dated September 30, 2021, the licensee noted that the containment pressure and temperature will change slightly due to the additional mass and energy released into the containment following a main steam line break (MSLB). SNC compared the revised parameters against the existing EQ test values for equipment located inside containment. The licensees review determined that there are no

impacts to the EQ of equipment inside and outside containment. The NRC staff understands that the physical plant modifications associated with this proposed LAR (i.e., replacement of the current MSIV actuator with a different type of actuator) will be incorporated using the 10 CFR 50.59 change process.

Based on the LAR and the licensees supplement dated January 13, 2022, the NRC staff also understands that the environmental conditions (e.g., temperature, pressure, radiation, chemical spray, humidity) outside containment during normal operation and under design bases events, will not be affected by the proposed changes and that the new MSIV actuators will be environmentally qualified in accordance with the Vogtle, Units 1 and 2, design and licensing basis for 10 CFR 50.49.

SNC did not provide sufficient information in its initial submittal on the impact of the proposed changes on the EQ of electrical equipment subject to the requirements of 10 CFR 50.49.

Accordingly, on November 15, 2021 (ADAMS Accession No. ML21321A376), the NRC staff requested the licensee to provide further details of its assessment on the impact the proposed changes have on the EQ of electrical equipment. One of the NRC staffs requests was for the licensee to address the following EQ parameters: temperature, pressure, radiation, humidity, and chemical spray and explain how electrical equipment will remain qualified, with margins maintained, ensuring continued compliance with 10 CFR 50.49.

By letter dated January 13, 2022, SNC provided an RAI response which noted that it evaluated a total of 24 equipment qualification data packages containing EQ components located inside containment to determine acceptability of the new MSLB temperature and pressure accident profiles. As a result of its review, the licensee determined that all EQ components inside containment impacted by the proposed change remain qualified to the new MSLB accident profiles with consideration of margin. Furthermore, the licensee noted that the proposed change does not impact the existing radiation, humidity, or chemical evaluations.

By letter dated November 15, 2021, the NRC staff requested the licensee to clarify whether any areas changed from mild to harsh environment, because of the proposed changes. In its letter dated January 13, 2022, SNCs response noted that no areas of the plant changed from a mild environment to a harsh environment because of the new MSLB accident profile. The NRC staff finds that this was likely the result of changes to environmental parameters being limited to inside containment, which is an area of the plant that was already classified as an EQ harsh environment.

By letter dated November 15, 2021, the NRC staff also requested the licensee to discuss the impact of the proposed change on non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1)(i)(A) through (C) of paragraph (b)(1) of 10 CFR 50.49 by the safety-related equipment. In its letter dated January 13, 2022, SNCs response noted that all non-safety-related equipment whose failure under postulated conditions could prevent the performance of a safety function are included on the Vogtle, Units 1 and 2, EQ Master List. The licensee stated that all EQ equipment on the EQ Master List located inside containment was evaluated for impacts from the new MSLB accident profile. According to SNC, this includes non-safety-related electrical equipment whose failure could impact a safety function. Based on the above, the NRC staff finds that no new electrical equipment located inside containment needs to be added to the licensees 10 CFR 50.49 EQ program.

Additionally, as environmental conditions outside containment during normal operation and

under design bases events will not be affected by the proposed changes, no new equipment located outside containment needs to be added to the licensees 10 CFR 50.49 EQ program.

Based on the above, the NRC staff finds that the proposed changes will have no adverse impact on the Vogtle, Units 1 and 2, EQ Program or its ability to continue to meet the requirements of 10 CFR 50.49. Therefore, the NRC staff finds that the EQ of electrical equipment inside and outside containment should remain bounded under the proposed changes, and is therefore, acceptable.

3.8 MSLB Analysis The SRP Section 10.3 states:

Assure that, in the event of a postulated break in a main steam line in a PWR

[pressurized-water reactor] plant, the design will preclude the blowdown of more than one steam generator, assuming a concurrent single active component failure. In this regard, all main steam shutoff valves downstream of the MSIVs, the turbine stop valves, and the control valves are considered to be functional.

An MSLB accident inside the containment occurring concurrently with a single failure of an MSIV is alleviated by the closing of the second MSIV in the same MSL. The proposed design change removes the automatic function from the outboard MSIV, so that only the inboard automatic MSIV will remain in each MSL.

The removal of the automatic operating function from the outboard MSIV in each of the MSLs has a potentially adverse impact on the MSLB containment pressure and temperature response analysis. The impact is because of the reverse flow from the steam header into the containment due to a single failure of the inboard MSIVs to isolate, when required, resulting in the blowdown of a SG following a MSLB. The blowdown effects are limited to one SG as the downstream turbine stop valves are assumed to close as expected. In the current configuration, this did not need to be considered as there was no credible single failure that could cause the reverse flow to occur. Based on the proposed TS change, the licensee has revised its analysis of MSLB mass and energy release into the containment and the associated containment pressure and temperature response analysis.

By letter dated January 13, 2022, SNC provided the following information concerning MSLB reanalysis: (a) methodology used, (b) assumptions, (c) changes in containment heat sinks, (d) changes in analysis design inputs, and (e) the sequence of events on which the analysis is based.

3.8.1 Methodology As stated in Vogtle, Units 1 and 2, UFSAR, Section 6.2.1.7, the licensee used the NRC-approved LOFTRAN methodology described in WCAP-8822-P/8860-NP (ADAMS Accession Nos. ML012340064 and ML19292G053) for mass and energy (M&E) release analysis, and the NRC-accepted COCO methodology described in WCAP-8326 (ADAMS Accession No. ML20301A596) for containment pressure and temperature response.

3.8.2 Analysis Assumptions SNC stated that the analysis used the same assumptions documented in UFSAR Table 6.2.1-2.

In addition, no changes were made in the system failure mode and effects analysis documented in UFSAR Table 6.2.2-3.

3.8.3 Containment Heat Sinks SNC stated that no changes were made to the modeling of thermophysical properties of containment heat sinks documented in UFSAR Table 6.2.1-5 for the COCO containment evaluation model.

3.8.4 Design Inputs The UFSAR Table 6.2.1-64 lists the design inputs used in the current analysis. By letter dated January 13, 2022, SNC stated that all cases listed in this table are reanalyzed, and changes made in cases 13, 14, 15, and 16 were provided in the letter. Cases 13 and 16 are the most limiting for containment temperature and pressure response, respectively. The significant change in the inputs that affects the results is in the un-isolatable MSL volume, which increased from 470 cubic feet (ft3) to 9961 ft3.

3.8.5 Results The updated analysis results show a slight increase in peak calculated containment temperature from 303 degrees Fahrenheit (°F) to 309°F and peak pressure increase from 36.5 pounds per square inch gauge (psig) to 39 psig. The revised results are less than the containment design pressure and containment atmosphere design temperature of 52 psig and 381°F, respectively, as documented in UFSAR Table 6.2.1-1.

The NRC staff reviewed SNCs MSLB re-analysis and the NRC staff finds:

The licensee used NRC-approved or accepted methodologies for the containment pressure and temperature response following a MSLB inside the containment.

The analysis assumptions, inputs, containment heat sink data, and sequence of events used are acceptable.

The analysis results are bounded by the containment design pressure and temperature of 52 psig and 381°F, respectively, documented in UFSAR Table 6.2.1-1.

The analysis based on the proposed change meets 10 CFR Part 50, Appendix A, GDC 16, because the calculated peak containment peak pressure and temperature due to MSLB inside the containment are bounded by the containment design pressure and temperature.

The containment integrity is maintained under the most severe secondary system pipe rupture, thus precluding the release of radioactivity to the environment.

Based on the above, the NRC staff finds that the design basis change from two MSIVs per steam line to one MSIV per steam line is acceptable with respect to the impact to the MSLB analyses.

3.9 Description of Changes for the MSIV Actuator In the letter dated September 30, 2021, SNC stated:

The physical plant change that is to occur to support the design basis change is the replacement of the current MSIV actuator with a different type of actuator.

This physical plant change has been evaluated against the criteria of 10 CFR 50.59 and determined to not require prior NRC approval.

The inboard MSIV is not changed as part of the proposed design change. The existing valve remains in the main steam line and remains capable of providing positive sealing in either direction. It remains capable of closing against full steam flow. No change to the design characteristics of this valve have occurred.

The outboard valve is not changed except for removal of the actuation signal and depressurization of the nitrogen accumulator. The outboard valve remains capable of closure and sealing against steam flow from either direction. It can be closed manually following operator action to charge the nitrogen accumulator for the existing actuator.

SNC proposes to change the licensing basis to eliminate the requirement for the automatic closure function from the outboard MSIV in each MSL. The change to the plant design will be implemented after approval of this LAR. Both MSIVs will remain in the MSL in their current location. The inboard MSIV will have a new dual train actuator installed with redundant actuation signals and redundant power supplies. The existing actuator will remain on the outboard valve (former MSIV) and will be capable of manual valve closure. Automatic closure of the outboard valve will be precluded by removal of the automatic actuation signal and depressurization of the nitrogen accumulator. The outboard valve will still be capable of manual closure using the existing actuator upon charging of the nitrogen accumulator.

The existing MSIV bypass valves remain in place around the remaining MSIV and the outboard valve. Each bypass valve retains a closure signal by either SLI-A or SLI-B and are capable of closing within the time assumed in the safety analysis. The design basis of the bypass valves is not changed in any way by the change to the operation of the MSIV.

As described in UFSAR Section 6.2.4.2.1, the containment penetrations associated with the secondary side of the steam generators are not subject to GDC 57. The valves associated with these penetrations do not receive a containment isolation signal and are not credited with effecting containment isolation in the safety analyses. The barriers against fission product release to the environment are the SG tubes and the piping associated with the SGs. These penetrations do have valves that are capable of remote manual operation and can serve to isolate these penetrations.

The design change from two automatic MSIVs per steam line to one automatic MSIV per steam line changes the result of a single failure analysis. The current design basis assumes a failure of one MSIV in one of the steam lines to close when required and accommodates a single failure without the need to assume operation of any downstream valves. With the removal of the automatic function from the current outboard MSIV, only one automatic MSIV will remain in each steam line. Therefore, a single failure of one of the MSIVs to isolate when required results

in the blowdown of a SG following a MSLB. See Section 3.10 below for the evaluation of a single MSIV design.

3.9.1 MSLB Containment Pressure and Temperature Response Removal of the function of one MSIV per steam line has an adverse impact on the containment pressure and temperature analyses due to allowing for reverse flow from the steam header into containment. This phenomenon did not need to be considered in a two-MSIV-per-steam-line configuration as there was no credible single failure that could cause this to occur.

The current AOR assumes simultaneous, dual single failures. These include a loss of one emergency diesel generator (EDG) (equates to a loss of one train of containment sprays and one train of fan coolers), and a failure of one feedwater (FW) isolation valve, consistent with UFSAR Section 6.2.1.4.6. While not required, assuming dual single failures did not have a significant impact on the containment pressure and temperature analysis results and allowed for the number of cases documented in the UFSAR to remain at a minimum (16 case break spectrum, UFSAR Table 6.2.1-65).

Consistent with the current AOR approach, the MSIV single failure is included along with the EDG and main FW isolation valve failures. This approach is conservative and bounding as it assumes three single failures occur simultaneously in the event. The updated analysis results show an increase in peak calculated containment temperature (303°F to 309°F) and pressure (36.5 psig to 39 psig).

The NRC staff reviewed the impacted safety analyses from two MSIVs per steam line to one MSIV per steam line. Due to the conservative approach of multiple simultaneous single failures in the containment pressure and temperature analysis, a small increase in peak containment temperature and pressure occurs. These increases result in containment pressure and temperature magnitudes less than the containment design limits of 52 psig and 381°F. As a result, the NRC staff finds that the design basis change from two MSIVs per steam line to one MSIV per steam line is acceptable with respect to the impact to the safety analyses.

3.10 Evaluation of Single MSIV Design SRP 10.3, Main Steam Supply System, was reviewed to ensure appropriate aspects of the change from two MSIVs to one MSIV per main steam line were properly considered. SRP 10.3 provides acceptance criteria to meet the relevant requirements of the NRCs regulations identified in this SRP section. The licensee provided an evaluation for each acceptance criteria in its LAR, and the NRC staff determined that the applicable acceptance criteria have been satisfied, as described below.

3.10.1 GDC 2 The NRC staff considered compliance with GDC 2, as it relates to safety-related portions of the system being capable of withstanding the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, and floods.

Vogtle, Units 1 and 2, conform with RG 1.29, Rev. 3 for the main steam system. With regard to regulatory position C.1, each nuclear steam supply system component important to safety is classified as Safety Class 1, 2, or 3; these classes are qualified to remain functional in the event of the safe shutdown earthquake. The licensee listed the following components as being

designed in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Class 2, and Seismic Category 1 requirements; these components are considered the safety-related portion of the system.

1.

All piping and valves from the steam generators up to and including the pipe restraints provided downstream of each outboard MSIV to maintain piping loads upstream of the restraint in accordance with Branch Technical Position MEB 3-1.

This safety related piping is designed to meet the no-break zone criteria of Nuclear Regulatory Commission Branch Technical Position MEB 3-1 so that the piping failures need not be postulated. The portion of the system designated as safety-related is not changed due to this design change.

2.

Branch lines from the above portions of the main steam lines up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic/remote manual closure during all modes of normal reactor operation.

The NRC staff reviewed the licensees proposed design and finds that the main steam system and MSIVs continue to meet the requirements of GDC 2.

3.10.2 GDC 4 The NRC staff considered compliance with GDC 4, with respect to safety-related portions of the system being capable of withstanding the effects of external missiles and internally generated missiles, pipe whip, and jet impingement forces associated with pipe breaks.

Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs. The fluid discharge for MSLBs downstream of the MSIVs changes from an initial discharge followed by isolation to an initial discharge followed by continued blowdown of one SG. The UFSAR describes that high energy line break (HELB) protection outside of containment against the dynamic effects of pipe failures is provided in the form of physical separation of systems and components, barriers, equipment shields, and pipe whip restraints.

The NRC staff reviewed the licensees design and finds that the protective mechanisms for the main steam system should not be adversely affected by a longer duration discharge, and therefore, the main steam system continues to meet the requirements of GDC 4.

3.10.3 GDC 5 The NRC staff considered compliance with GDC 5, as it relates to the capability of shared systems and components important to safety to perform required safety functions.

Vogtle, Units 1 and 2, is a two-unit plant with the following common safety-related structures:

A. Control building B. Auxiliary building C. Fuel handling building Within these buildings are shared spaces, such as the control room, which contain physically separated safety-related equipment. Safety-related systems are not shared, except for the fuel

handling building post-accident exhaust system. Common heating, ventilation, and air conditioning (HVAC) system ducting headers are used in some instances for redundant HVAC units.

The NRC staff reviewed the licensees design and finds that the main steam system is not shared between Units and is not affected by the requirements of GDC 5.

3.10.4 GDC 34 The NRC staff considered compliance with GDC 34, as it relates to the system function of transferring residual and sensible heat from the reactor system in indirect-cycle plants.

The residual heat removal (RHR) system, in conjunction with the steam and power conversion system, is designed to transfer the fission product decay heat and other residual heat from the reactor core at a rate which keeps the fuel within acceptable limits. The RHR system functions when temperature and pressure are below approximately 350°F and 400 psig, respectively.

Redundancy of the RHR system is provided by two residual heat removal pumps (located in separate flood-proof compartments, with means available for draining and monitoring leakage),

two heat exchangers, and associated piping, cabling, and electric power sources. The RHR system can operate on either the onsite or offsite electrical power system. Redundancy of heat removal at temperatures above approximately 350°F is provided by the four steam generators, four atmospheric relief valves, and the auxiliary feedwater system.

A review was performed of the main steam system to consider the isolation capability downstream of the MSIVs. In those cases when an MSIV fails to close, isolation of the main steam system must be performed downstream of the MSIVs. This can involve closure of several valves either automatically, remotely from the control room, or locally depending on the system design. These valves may be considered non-safety class in some cases. Potential leakage through certain valves was identified and considered in the SGTR analysis.

The main steam safety valves are the safety-related residual heat removal path while above RHR entry conditions. The staff reviewed the licensees proposed design and found that it does not affect the main steam safety valves because they are upstream of the MSIVs. Therefore, the staff finds that the main steam system with one MSIV per main steam line continues to meet the requirements of GDC 34.

3.10.5 10 CFR 50.63 The NRC staff considered compliance with 10 CFR 50.63, as it relates to the ability of a plant to withstand for a specified duration and then recover from an SBO.

Vogtle, Units 1 and 2, follow methods in Regulatory Guide 1.155, Revision 0. Main steam isolation can be achieved by closing the MSIVs and closure of bypass valves. The MSIVs close on loss of control or actuation power. Therefore, the MSIVs will close following a station blackout. The bypass valves are not changed from the existing design.

The NRC staff reviewed the licensees proposed design and finds that the main steam system continues to meet RG 1.155, Revision 0, and, therefore, meets 10 CFR 50.63.

3.10.6 RG 1.29 The NRC staff considered methods acceptable to NRC in RG 1.29, Positions C.1.a, C.1.e, C.1.f, C.2 and C.3, as it relates to the seismic design classification of system components.

As described above under GDC 2, Vogtle, Units 1 and 2, meet RG 1.29, Rev. 3 for the main steam system and, therefore, the licensees proposed design is acceptable as it relates to the seismic design classification of the main steam system components.

3.10.7 RG 1.117 The NRC staff considered methods acceptable to NRC in RG 1.117, Appendix Position 2 and 4, as it relates to the protection of SSCs important to safety from the effects of tornado missiles.

Vogtle, Units 1 and 2, meet RG 1.117, Revision 1, except for the nuclear service cooling water tower fans, main steam safety valve exhausts, atmospheric relief valve (loop 2), atmospheric relief valve exhaust stacks, turbine-driven auxiliary feedwater pump exhaust, and condensate storage tank vents which are not missile protected.

The main steam system and MSIVs are otherwise protected. This level of protection has not changed from the current design basis. Therefore, the NRC staff finds that the licensees proposed design is acceptable as it relates to tornado missile protection.

3.10.8 SECY-93-087 SECY 93-087 applies to boiling-water reactor plants and, therefore, is not applicable to Vogtle, Units 1 and 2, since both units are pressurized-water reactor plants.

3.11 Evaluation of Inboard MSIV Actuators - System In its letter dated September 30, 2021, SNC stated:

The physical plant change that is to occur to support the design basis change is the replacement of the current MSIV actuator with a different type of actuator.

This physical plant change has been evaluated against the criteria of 10 CFR 50.59 and determined to not require prior NRC approval.

By letter dated September 30, 2021, SNC requested the reduction from two MSIV systems per steam line to one MSIV per steam line. With the proposed design change, there will no longer be MSIV systems, and each MSL will have one MSIV with a dual-train actuator.

In addition, the two bypass valves remain in each MSL, and each is closed by a different train of the SLI signal.

In the SNC letter dated September 30, 2021, SNC stated:

SNC proposes to change the licensing basis to eliminate the requirement for the automatic closure function from the outboard MSIV in each main steam line. The change to the plant design will be implemented upon approval of this licensing change. Both MSIVs will remain in the main steam line, in their current location.

The inboard MSIV will have a new dual train actuator installed with redundant

actuation signals and redundant power supplies. The existing actuator will remain on the outboard valve (former MSIV) and will be capable of manual valve closure. Automatic closure of the outboard valve will be precluded by removal of the automatic actuation signal and depressurization of the nitrogen accumulator.

The outboard valve will still be capable of manual closure using the existing actuator upon charging of the nitrogen accumulator. The MSIV bypass valves remain in place with their existing closure signals. The non-safety related valves downstream of the outboard valve are not altered by this change.

MSIV actuators are considered integral to the MSIV, so the inoperability of an MSIV actuator constitutes inoperability of the MSIV and appropriate MSIV Conditions are required to be entered to address the inoperability of an MSIV. The licensee proposes to install a design change, a new style dual train actuator is installed on the inboard MSIV. The dual train actuator allows the inboard MSIV to close with a single failure of one train of the actuator. In addition, the inboard MSIV dual train actuator receives closure signals from both trains of SLI, with each train of the actuator receiving a train signal. The final design of the inboard MSIV closure for each steam line will consist of a single inboard MSIV with a dual train actuator and two bypass valves. The bypass valves will retain a single train of SLI closure signal to each valve.

Currently, Vogtle, Units 1 and 2, TS 3.7.2 does not specifically address or reflect two dual train actuators for one inboard MSIV, and inoperability of one of the two actuator trains associated with an inboard MSIV will not by itself make the valve incapable of closing since the remaining operable actuator train can alone close the valve on demand. Therefore, the licensee is proposing to incorporate appropriate Conditions and Required Actions to address inoperable inboard MSIV actuator trains in TS 3.7.2.

The proposed revised text specifically identifies the MSIV actuator trains as equipment required to be OPERABLE for safe operation of the facility. The MSIV actuator trains are required to close the MSIVs, when needed. The NRC staff finds this design change acceptable, because the dual train actuator design for each MSIV ensures that a single inoperable actuator train for any MSIV would not prevent the affected MSIVs from closing on demand.

3.11.1 Evaluation of Inboard MSIV Actuators - Electrical The NRC staff reviewed the licensees submittal dated September 30, 2021, response to RAI questions dated January 13, 2022, and the latest UFSAR to determine the impact of the proposed change on the electrical equipment. In the LAR, SNC described that for each MSL that there are two MSIVs in series, one closer to the SGs (inboard) and one closer to the turbine (outboard). The outboard MSIVs will remain in place with actuators set up for manual control only. The inboard MSIVs will have the actuators replaced with a different type and will now have redundant power to meet the single failure criterion. Previously, power for the inboard and outboard MSIVs was supplied by separate trains.

By letter dated September 30, 2021, the licensee described how the current MSIV configuration relies on a single train per MSIV with single failure met by redundant MSIVs controlled by separate trains. The modified system will keep the outboard valve and actuator, but it will be modified to respond only to manual control. The inboard MSIVs will use a dual train actuator.

The NRC staff reviewed the LAR and determined that the modified electrical design will continue to meet the single failure criteria of IEEE 279-1971, because each train can still independently accomplish the safety function by closing the MSIV on an isolation signal and power from either train.

By letter dated January 13, 2022, SNC stated, in part, that the approved design calculations for the first Vogtle, Unit 2, design change have evaluated the impacts of the MSIV actuator load change on the electrical distribution system and the results indicate the load additions are within the capacities of their respective sources and as a result will not impede the operation of other critical loads. The NRC staff determined that because the loading changes will not impact operation of critical loads and that battery margin is preserved, there is reasonable assurance that the proposed change will continue to meet the general functional criteria of IEEE 279-1971.

SNC further stated, in the letter dated January 13, 2022, that calculations indicate sufficient available margins in the batteries and chargers during LOOP/LOCA and SBO and as a result, the load additions are acceptable and will not impede operation of other critical loads. The NRC staff determined that because the loading changes will not impact operation and battery margin is preserved, there is reasonable assurance that the proposed change will continue to meet the criteria of 10 CFR 50.63.

The NRC staff determined that because the proposed change continues to meet the general functional and single failure criteria of IEEE 279-1971, that it continues to meet the requirements of GDC 17.

Based on the above, the NRC staff finds that the proposed changes will have no adverse impact on the Vogtle, Units 1 and 2, or its ability to continue to meet the requirements of 10 CFR 50.49, GDC 17, 10 CFR 50.55a(h), and 10 CFR 50.63.

3.12 Evaluation of MSIV Bypass Valves The MSIV bypass valves are normally open during plant operation, while the MSIVs are also open, to minimize condensation in the steam line and to support maintenance and testing at power. The MSIV bypass valves may also be opened when the MSIVs are closed for warming of the steam lines and equalizing steam pressure across the MSIVs. The normally open bypass valves could allow a potential release flow path to exist following a postulated accident scenario.

Therefore, the bypass valves have been subject to the same requirements for isolation as the MSIVs.

Currently, Vogtle, Units 1 and 2, TS 3.7.2 requires two MSIV systems to be operable for each steam line. As defined in the TS Bases, a MSIV system consists of an MSIV and its associated bypass valve; however, Vogtle, Units 1 and 2, current design has two MSIVs and two bypass valves in each steam line. With the licensees proposed design change, a separate TS Condition and Required Actions is added for the main steam isolation bypass valves.

With the proposed change, both existing bypass valves will remain operable as currently described in the TS. The MSIV bypass valves will continue to each have a separate train closure signal (SLI-A or SLI-B), ensuring that one of the two bypass valves will close following a single failure.

3.13 Evaluation of TS Changes 3.13.1 Proposed LCO 3.7.2 The proposed changes to LCO 3.7.2 are marked in Section 3.1.1 of this safety evaluation.

SNC proposed to add dual train actuators and the bypass valves to the LCO, Conditions and SRs. The LCO is revised to account for the modifications that will consist of a single MSIV with a dual train actuator and two bypass valves for each steam line. Both bypass valves are required to be in the LCO for each steam line. The NRC staff finds the proposed change acceptable, because the revised LCO specifically includes the MSIV, the dual train actuators and both bypass valves in each steam line. This structure allows for the addition of actuator Conditions and a bypass valve Condition. The LCO Applicability continues to address the MSIV and bypass valves, which specifies the MSIV and the bypass valves separately. This ensures that all required valves are closed, if needed, for the exception to the Applicability.

3.13.2 Proposed Deletion of NOTE The proposed deletion of a NOTE for the ACTION statements is marked in Section 3.1.2 of this safety evaluation. The NOTE allowing separate Condition entry for all Actions is removed from its current location, as it is no longer appropriate to allow separate Condition entry for each steam line with the change to the LCO from two MSIV systems to one MSIV and associated actuators and bypass valves per steam line. The NOTE is added to Conditions H and I. The NRC finds the proposed change acceptable because the NOTE is no longer needed in the ACTIONS for TS 3.7.2.

3.13.3 Proposed Conditions A through E The licensee proposed new Conditions A through E as marked in Section 3.1.3 of this safety evaluation.

For Condition A, with only a single actuator train inoperable on one MSIV, the licensee proposed a Completion Time of 7 days for Required Action A.1. With one actuator train inoperable, and because of the dual train actuator design, the affected MSIV would still be capable of closing on demand (assuming no additional failures) via the remaining Operable actuator. The proposed 7-day Completion Time takes into account the design redundancy, reasonable time for repairs, and the low probability of a design basis accident occurring during this period.

For Condition B, with an inoperable actuator train on one MSIV and one inoperable actuator train on another MSIV, such that the actuator trains are not in the same train (A or B), the license proposed a Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for Required Action B.1. The dual train actuator design ensures that with only one actuator train inoperable on each of the affected MSIVs, each MSIV would still be capable of closing on demand, assuming no additional failures.

Compared to Condition A, a shorter allowed outage time is appropriate for Condition B since with an actuator train inoperable on each of two MSIVs, there is an increased likelihood that an additional failure (such as the failure of an actuation logic train) would cause an MSIV to fail to close.

For Condition C, with one inoperable actuator train on one MSIV and one inoperable actuator train on another MSIV, but with both inoperable actuator trains in the same train, the licensee proposed a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Required Action C.1. The dual train actuator design for each MSIV ensures that a single inoperable actuator train for any MSIV would not prevent the affected MSIVs from closing on demand. The licensee proposed that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is reasonable and conservative since only one actuator train per MSIV is permitted to be inoperable (for two MSIVs), so that the remaining operable actuator train on each affected MSIV remains capable of effecting valve closure on demand (assuming no additional failures). A Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is also considered conservative with respect to the low probability of an event occurring during such an interval that would demand MSIV closure. However, compared to the Required Action for Condition B above, a shorter Completion Time for Condition C is appropriate since with two actuator trains inoperable in the same train, an additional failure such as the failure of actuation logic in the other train could cause both affected MSIVs to fail to close on demand.

For Condition D, with two (both) actuator trains on one MSIV inoperable, the licensee proposed an immediate Completion Time for declaring the MISV inoperable for Required Action D.1. The immediate Completion Time is appropriate since having both actuator trains inoperable would constitute a condition that renders the affected MSIV incapable of closing on demand. If no other MSIV or MSIV actuation train inoperability existed at the time of this Condition, this would result in new Condition F or I (depending on the applicable plant Mode) being immediately entered for a single inoperable MSIV.

For Condition E, with three or more MSIV actuator trains inoperable, OR (after entering Conditions A, B or C) it is determined that the Required Action and Completion Time of Condition A, B, or C cannot be met, the licensee proposed an immediate Completion Time for declaring each affected MSIV inoperable for Required Action E.1. Declaring only a single MSIV inoperable (due to the Required Action and Completion Time of Condition A not being met), would result in entry into Condition F (for MODE 1) or Condition I (for MODE 2 or 3).

Declaring more than one MSIV inoperable would result in entry into LCO 3.0.3 (for MODE 1) or Condition I (for MODE 2 or 3), since (for the former) there is no Condition under TS 3.7.2 that addresses having more than one MSIV inoperable during MODE 1. When only the actuator trains for affected MSIVs are inoperable (and not the valves themselves), the Conditions and Required Actions for the inoperable MSIV actuator trains should be entered first, and then if those Required Actions cannot be met, the affected MSIVs should be declared inoperable so that the Conditions and Required Actions for the inoperable valves are then entered. Required Action E.1 ensures the affected MSIV(s) is promptly declared inoperable.

For the other part of Condition E, i.e., for the condition when three or more actuator trains are inoperable, such a condition could involve two inoperable actuator trains on one MSIV and one inoperable actuator train on another MSIV, or an inoperable actuator train on each of three MSIVs. In each case, the inoperable actuator trains could all be in the same train or be staggered among the two trains. In the former case, a single assumed failure such as an instrument logic train failure could cause one or two MSIVs to fail to close on demand. In the latter case, such a single failure could cause all three MSIVs to fail to close on demand. The conditions addressed by Condition E would constitute an inoperability that exceeds the scope of any of the conditions addressed by Conditions A, B, or C, and immediately declaring the MSIVs inoperable is appropriate.

The NRC staff found the proposed changes to Conditions A through E acceptable because the proposed changes are based on a hierarchy of Conditions such that shorter CTs would be specified for increasingly degraded conditions consistent with STS format.

3.13.4 Current Condition A - Proposed Condition F The licensee proposed to change the current Condition A to reflect the change from two MSIV systems per steam line to one MSIV per steam line, and to re-label Condition A as Condition F.

In addition, the license proposed to delete the allowance for more than one MSL to have an inoperable MSIV system. SNC proposed the CT for restoration of the inoperable MSIV to be changed from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee proposed to eliminate the RICT. The marked-up changes are given in Section 3.1.4 of this safety evaluation.

The NRC staff finds the proposed change acceptable because the proposed Condition reflects the plants proposed design and the CT for Condition F is based on a hierarchy of Conditions consistent with STS format.

3.13.5 Current Condition B The current Condition B is proposed to be deleted as marked in Section 3.1.5 of this safety evaluation. With only one MSIV per steam line, this Condition no longer applies. The NRC staff finds this proposed change acceptable because the Condition is no longer applicable.

3.13.6 Current Condition C - Proposed Condition G The current Condition C is proposed to be re-labeled to Condition G and now applies only to a failure to meet the Required Action and CT of Condition F (as opposed to the current Condition C that applies upon failure to meet the Required Action and CT of Conditions A or B). The reference to previous Condition B is removed. The marked-up changes are given in Section 3.1.6 of this safety evaluation.

The NRC staff finds the proposed change acceptable because the proposed Condition G is based on a hierarchy of Conditions consistent with STS format.

3.13.7 Current Condition D The current Condition D is proposed to be deleted as marked in Section 3.1.7 of this safety evaluation. This Condition no longer applies with one MSIV per steam line. The NRC staff finds this proposed change acceptable because the Condition is no longer applicable.

3.13.8 Proposed Condition H As marked in Section 3.1.8 of this safety evaluation, SNC proposed a new Condition H to address an inoperable bypass valve. A separate Condition entry would be allowed for each bypass valve because it operates independently of the other bypass valves.

New Condition H addresses one or more bypass valves inoperable and Required Actions H.1 and H.2 require the inoperable bypass valves to be closed or isolated within 7 days and verified closed or isolated once per seven days.

The proposed LCO and Applicability identify that bypass valves are associated with each MSIV.

The proposed changes to the LCO and Applicability more clearly convey the requirements and provisions on a for each main steam line basis and provide clear exceptions to the Applicability which are permitted on the basis that isolation or isolation capability is still ensured when the exception is involved.

Under the proposed change, the bypass valves are required to be Operable in MODES 1, 2, and 3. The bypass valves are considered Operable when their isolation times are within limits and they are capable of closing on an isolation actuation signal. All bypass valves can be closed at power. Consistent with the Westinghouse STS, exceptions to the Applicability are allowed for the MSIVs in MODES 2 and 3. In MODES 2 and 3, with all MSIVs closed, the MSIVs are assured of performing their specified safety functions. Likewise, with one bypass valve in each steam line closed, the bypass valves are assured of performing their specified safety function. With the assurance that the specified safety function is being met, it is acceptable to exempt the MSIVs and bypass valves from the Applicability of TS 3.7.2 under such conditions. Therefore, in MODES 2, and 3, exceptions to the Applicability for TS 3.7.2 are allowed for the bypass valves when they are assured of performing their specified safety function.

With one or more bypass valves inoperable, the valve(s) must be closed or isolated within 7 days and verified closed or isolated once per 7 days. The bypass valve can be isolated by closing the redundant bypass valve in the same main steam line or valves in the bypass line downstream of the bypass valves, which ensures the bypass line flow is isolated supporting the assumptions in the safety analysis. The NRC staff finds the licensees proposed 7-day CT to be acceptable, because it takes into consideration the low probability of an accident occurring during this time that would require a closure or isolation of the bypass valve and less significant consequences from a postulated accident following failure of a bypass valve to isolate. For an inoperable bypass valve that cannot be restored to Operable status within the specified Completion Time, but is closed or isolated, the inoperable bypass valve must be verified on a periodic basis to be closed or isolated. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7-day CT is consistent with the CT for an inoperable, but closed MSIV in MODE 2 or 3.

3.13.9 Current Condition E - Proposed Condition I SNC proposed to re-label current Condition E to Condition I as marked in Section 3.1.9 of this safety evaluation. The licensee proposed to add a NOTE for separate Condition entry to Condition I. The Condition is changed to reflect the change from two MSIV systems per steam line to one MSIV per steam line. The licensee proposed Required Action I.1 be added to close or isolate the inoperable MSIV, which can be accomplished by closing the outboard valve in the same MSL. This ensures the MSL flow is isolated supporting the assumptions in the safety analysis. SNC proposed to change the CT from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee proposed to modify Condition I.2 to reflect the change from two MSIV systems per steam line to one MSIV per steam line. It requires verification that the MSIV is closed or isolated.

The NRC staff finds the proposed change acceptable, because the proposed Condition reflects the plants proposed design change, and the CT for Condition I is based on a hierarchy of Conditions consistent with STS format.

3.13.10 Current Condition F - Proposed Condition J The licensee proposed to re-label current Condition F to Condition J. As marked in Section 3.1.10 of this safety evaluation, SNC proposed that the reference to the current Conditions D and E be changed to proposed Conditions H and I. Proposed Condition I is a relabeled version of Current Condition E, while current Condition D is proposed for deletion and Condition H is new.

The NRC staff finds the proposed change acceptable because the proposed Condition J is based on a hierarchy of Conditions consistent with STS format.

3.13.11 Proposed Changes to SR 3.7.2.1 SNC proposed to change SR 3.7.2.1 to reflect the change from an MSIV system to an MSIV and bypass valve as marked in Section 3.1.11 of this safety evaluation for consistency with separate Conditions for MSIVs and bypass valves.

The NRC staff finds the proposed change acceptable because the SR supports the separation of the MSIVs and bypass valves in the LCO and the addition of a separate bypass valve Condition.

3.13.12 Proposed SR 3.7.2.2 SNC proposed to add SR 3.7.2.2 to address the dual train actuators on the MSIVs as marked in Section 3.1.12 of this safety evaluation. The NRC staff finds the proposed SR to be acceptable, because the SR addresses the ability of each of the dual train actuators to close the MSIV as required per the LCO and includes verification that each dual train actuator is capable of independently closing its associated MSIV on an actual or simulated actuation signal. The frequency of actuator testing is acceptable because it is in accordance with the Surveillance Frequency Control Program, and the initial frequency in the SFCP is every 36 months, consistent with the current actuator testing cycle and the Vogtle, Units 1 and 2, refueling cycle.

3.13.13 Proposed TS Changes - Conclusion Based on the above, the NRC staff concludes that the TS, as modified by the proposed changes, would continue to meet the regulatory requirements of 10 CFR 50.36. In accordance with 10 CFR 50.36(c)(2)(i), when an LCO is not met, the licensee is required to shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met. The license amendment request would remove part of the permissible remedial actions from LCO 3.7.2, revise other remedial actions from LCO 3.7.2, and add remedial actions to LCO 3.7.2.

Under the Commissions regulations in 10 CFR 50.92 and 50.57, to issue the amended TSs, the Commission must be able to find, among other things, that the amended remedial actions provide reasonable assurance of public health and safety, and the NRC staff finds that the proposed remedial actions provide the requisite reasonable assurance. The NRC staff also concludes that the LCO was appropriately modified to reflect the applicants proposed design change and continues to establish the lowest functional capability or performance levels of equipment required for safe operation of the facility as required by 10 CFR 50.36(c)(2)(i).

The NRC staff reviewed the proposed changes to the TSs and determined that they meet the regulations for TSs in 10 CFR 50.36(b), because, as discussed above, proposed changes are based on a hierarchy of Conditions such that shorter CTs would be specified for increasingly degraded conditions consistent with STS format.

The NRC staff concludes that the proposed SRs assure that the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met and satisfy 10 CFR 50.36(c)(3).

Additionally, the NRC staff confirmed that the changes to the TSs were technically clear and consistent with customary terminology and format in accordance with Chapter 16 of the SRP.

The NRC staff concludes that the proposed TS changes satisfy 10 CFR 50.36, 10 CFR 50.49, 10 CFR 50.63, and other applicable standards. Based on the above, the NRC staff concludes that the proposed TS changes for Vogtle, Units 1 and 2, are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified on April 25, 2022. On April 25, 2022, the State of Georgia official replied that the official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration that was published in the Federal Register on December 28, 2021 (86 FR 73819), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: John G. Lamb, Steve Wyman, Nan Chien, Sean Meighan, Brian Lee, Ahsan Sallman, Matt McConnell, Hang Vu, Gurjendra Bedi, Nadim Khan, Andrea Russell, and Yueh-Li Li Date: July 5, 2022

ML22116A084 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DEX/EMIB/BC NRR/DSS/STSB/BC NAME JLamb KGoldstein ITseng VCusumano DATE 4/6/2022 05/03/2022 4/6/2022 4/25/2022 OFFICE NRR/DEX/ELTB/BC NRR/DSS/SCPB/BC NRR/DSS/SNSB/BC NRR/DRA/APLB/BC NAME JJohnston (WMorton for)

BWittick SKrepel JWhitman DATE 4/7/2022 4/11/2022 4/11/2022 3/17/2022 OFFICE NRR/DRA/ARCB/BC NRR/DEX/EEEB/BC OGC - NLO NRR/DORL/LPL2-1/BC NAME KHsueh WMorton MASpencer MMarkley DATE 4/11/2022 4/18/2022 6/23/2022 7/5/2022 OFFICE NRR/DORL/LPL2-1/PM NAME JLamb DATE 7/5/2022