ML23037A091

Safety Evaluation - Vogtle Unit 3 Amendment No. 190, Tech Spec Exceptions for In-Containment Refueling Water Storage Tank Operability Prior to Initial Criticality (Exigent Circumstances)

ML23037A091
Person / Time
Site:	Vogtle
Issue date:	02/08/2023
From:	Bill Gleaves NRC/NRR/VPOB
To:	Whitley B Southern Nuclear Operating Co
Shared Package
ML23037A082	List: ML23037A088 ML23037A089 ML23037A090 ML23037A091
References
Download: ML23037A091 (1)

Text

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 190 TO COMBINED LICENSE NO. NPF91 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 3 DOCKET NO. 52025

1.0 INTRODUCTION

By letter dated January 31, 2023 (Agencywide Documents Access and Management System Accession No. ML23031A359), Southern Nuclear Operating Company, Inc. (SNC or the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC) amend the Vogtle Electric Generating Plant (VEGP) Unit 3, Combined License (COL) No. NPF91. The License Amendment Request (LAR) 23004, Exigent License Amendment Request: Technical Specification Exceptions for In-containment Refueling Water Storage Tank Operability Prior to Initial Criticality, proposed changes to the VEGP, Unit 3, COL Appendix A, Technical Specifications (TS) to remove operability requirements for 1) the In-containment Refueling Water Storage Tank (IRWST) in TS operating Modes 5 and 6 prior to initial criticality, and 2) various automatic and manual actuation signals that initiate the IRWST, Automatic Depressurization System (ADS) Stage 4, and the Chemical and Volume Control System (CVS) prior to initial criticality. These are permanent changes to the VEGP Unit 3 TS and are temporary in the sense that they only apply prior to initial criticality, a one-time event.

SNC requested these TS changes in order to facilitate a repair to VEGP Unit 3 IRWST injection isolation valve 3PXS-V123A. To repair this valve, SNC may need to first drain down the coolant in the reactor coolant system (RCS). The proposed TS changes allow SNC to do this drain down without having to meet the current TS requirement to perform a total core offload, or defueling, of the reactor. The defueling, repair, and refueling of the reactor and associated activities are estimated to add over 16 days to the startup activities currently underway for Unit 3. Because of this impact on the critical path schedule to reactor startup previously discussed, SNC proposed this LAR to enable the Enclosure

2 repair without defueling and return to planned startup activities as rapidly as safely achievable.

The current TS requirements for the IRWST injection and recirculation flow paths in TS 3.5.7 and TS 3.5.8 are based, in part, on the need to remove decay heat following power operation of the reactor. In contrast, the VEGP Unit 3 reactor has a fresh core of fuel that has never undergone criticality or power operations.

In its LAR, the licensee requested that the NRC process the proposed amendment on an exigent basis. The LAR further stated that approval of the proposed amendment was requested within 8 days of the LAR submittal to allow the option to progress with IRWST injection isolation valve repair activities while minimizing the potential impact on the ongoing startup activities.

2.0 REGULATORY EVALUATION

The staff considered the following regulatory requirements in reviewing the LAR:

10 CFR 52.98(f) provides that any modification to, addition to, or deletion from the terms and conditions of a COL is a proposed license amendment. These activities involve a change to COL Appendix A TS information. Therefore, NRC approval in the form of a license amendment is required prior to making these plant-specific proposed changes.

10 CFR Part 52, Appendix D, VIII.C.6, states that after issuance of a license, Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90. 10 CFR 50.90 addresses the application for amendment of a license, including a combined license. The proposed LAR requires changes in the TS, and therefore an LAR is required to be submitted for NRC approval.

The regulation at 10 CFR 50.36(b) requires:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) surveillance requirements (SR); (4) design features; and (5) administrative controls. The rule does not delineate the specific requirements to be included in a plants TS. The regulation also states, in part, that [a]

summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

As stated in 10 CFR 50.36(c)(2)(i), the Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications.

3 10 CFR 50.91(a)(6) provides the conditions under which the Commission may find that an exigent situation may exist, the associated process for publishing a reasonable notice to the public in the area surrounding the licensees facility and consulting with the licensee on the media release and its geographic area of coverage, providing any public comments to the licensee, publishing a notice of issuance under 10 CFR 2.106, providing a hearing after issuance for a hearing request satisfying 10 CFR 2.309, requiring the licensee to explain the exigency and why it cannot be avoided, and what notice and comment procedures will be provided if the NRC determines that the licensee has failed to use its best efforts to make a timely application for the amendment in order to create the exigency and to take advantage of this procedure.

The NRC principal design requirements for nuclear power plants are the general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. VEGP Unit 3 was designed to satisfy the following GDC:

GDC 20, Protection system functions, requires that the protection system shall be designed (1) to automatically initiate the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded as a result of anticipated operational occurrences (AOOs), and (2) to sense accident conditions and initiate operation of systems and components important to safety.

GDC 34, Residual heat removal, requires that the plant design include a system to remove residual heat from the reactor core so specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

3.0 TECHNICAL EVALUATION

3.1 Description of Proposed Changes The proposed changes add Notes (and Footnotes) to modify the Modes 5 and 6 operability requirements for the IRWST injection and recirculation flow paths, the IRWST and ADS Stage 4 actuation instrumentation and controls, and the CVS automatic isolation function so that they would not be required to be operable in Modes 5 and 6 prior to initial criticality of the core.

Specifically, the licensee proposed revising the Applicability of the Limiting Condition for Operation (LCO) for the following LCOs:

1) LCO 3.3.8, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, Table 3.3.81 a) Function 14, RCS Wide Range Pressure - Low, b) Function 16, CMT [Core Makeup Tank] Level - Low 6, and c) Function 18, IRWST Lower Narrow Range Level - Low 3;

2) LCO 3.3.9, ESFAS Manual Initiation, Table 3.3.91 a) Function 7, ADS Stage 4 Actuation - Manual Initiation, b) Function 12, In-Containment Refueling Water Storage Tank (IRWST) Injection Line Valve Actuation - Manual Initiation, and c) Function 13, IRWST Containment Recirculation Valve Actuation - Manual Initiation;

3) LCO 3.3.10, ESFAS Reactor Coolant System (RCS) Hot Leg Level Instrumentation, Table 3.3.101

4 a) Function 1, Hot Leg Level - Low 4, and b) Function 2, Hot Leg Level - Low 2;

4) LCO 3.3.19, DAS Manual Controls, Table 3.3.19-1 a) Function 7, ADS stage 4 valves, b) Function 8, IRWST injection squib valves, and c) Function 9, Containment recirculation valves;

5) LCO 3.3.20, ADS and IRWST Injection Blocking Device, Table 3.3.20-1 a) Function 2, ADS and IRWST Injection Block Switches for Manual Unblocking;

6) LCO 3.5.7, IRWST - Shutdown, MODE 5;

7) LCO 3.5.8, IRWST - Shutdown, MODE 6; and

8) LCO 3.3.16, ESFAS Actuation Logic - Shutdown.

For Items 1 through 5 listed above, the Mode 5 and 6 Applicability for each of the instrument channel actuation functions would be modified by a footnote that states, For Unit 3 only, not required to be OPERABLE prior to initial criticality. For Items 6 and 7 listed above, the licensees proposed changes would modify the Applicability for Modes 5 and 6 for the IRWST by adding a note to LCOs 3.5.7 and 3.5.8 that states, For Unit 3 only, not required to be OPERABLE prior to initial criticality. Similarly, for LCO 3.3.16 (Item 8), the proposed changes would rename the current NOTE to be NOTES, number the current note to be 1. and add a note 2. that states, For Unit 3 only, ESF Actuation Function for ADS Stage 4 flow paths, In-Containment Refueling Water Storage Tank injection and recirculation flow paths, and CVS letdown isolation valves, not required to be OPERABLE prior to initial criticality.

The proposed changes would allow the licensee to operate in Modes 5 and 6 with the affected LCOs not met (i.e., the affected equipment and instrument channel actuation functions inoperable) prior to initial criticality in order to facilitate repair of the 3PXS-V123A valve without having to defuel the reactor.

Identification of Error in Unit 3 Technical Specification page 3.3.88 In reviewing the licensees proposed changes, the NRC identified typographical errors for the names of Functions 18, 20, and 21 in TS Section 3.3.8 in previous NRC-issued license amendments. The function name changes had previously been requested by SNC and approved by NRC in Unit 3 Amendment No. 100 (Ref. 2). However, follow-on Unit 3 Amendment No. 118 (Ref. 3) and Amendment No. 148 (Ref. 5) were issued with a change page containing the incorrect Function names. This issue is addressed in Section 3.2.

3.2 Evaluation of Changes The staff evaluated the proposed changes to determine if:

1. The licensees proposed changes are compliant with all applicable regulatory requirements, and

2. Operating the plant in accordance with the proposed changes provides reasonable assurance of adequate protection of public health and safety.

As stated in 10 CFR 50.36, The technical specifications will be derived from the analyses and evaluation included in the safety analysis report. In addition, the LCO defines the lowest

5 functional capability or performance levels of equipment required for safe operation of the facility. The staff reviewed the licensees Updated Final Safety Analysis Report (UFSAR) to determine if the safety analyses supported the licensees assertion that, prior to initial criticality of the core, the IRWST injection and recirculation flow paths, the IRWST and ADS Stage 4 actuation instrumentation and controls, and the CVS automatic isolation function affected by the proposed change do not perform a safety function and, therefore, do not need to be operable in Modes 5 and 6 prior to initial criticality of the fuel in the core.

The IRWST provides multiple safety functions. Specifically, during non-LOCA (loss of coolant accident) events, the IRWST serves as the initial heat sink for the Passive Residual Heat Removal (PRHR) Heat Exchanger (PRHR HX) if the normal residual heat removal system (RNS) is unavailable during reactor cooldown to MODE 4. During LOCAs or automatic depressurization events, it provides a source of low pressure safety injection. If the RNS fails during mid-loop operation with the ADS valves open, it can provide core cooling until the containment is flooded sufficiently for recirculation cooling. Based on these safety functions, the staffs evaluation considered whether operation with the requested TS changes would provide reasonable assurance that:

1. core cooling would not be required prior to initial criticality,

2. a loss of coolant inventory while performing the valve repair in Modes 5 and 6 would not require passive coolant injection, and

3. the core would remain covered with borated water and that adequate shutdown margin (SDM) would be maintained.

The primary technical justification provided by the licensee in support of the proposed TS amendments is that prior to the initial criticality, with unirradiated assemblies, there is no decay heat present in the core. Since the core fuel is unirradiated, there are no fission products in the core and there would not be any radiological consequences if the IRWST injection and recirculation flow paths are not available for core cooling. The NRC staff agrees with this technical justification. The NRC staff focused its evaluation on the controls available to ensure that the core remains submerged with borated water as required by TS 3.1.1 in Mode 5 and to prevent inadvertent dilution of boron in the core.

The NRC staffs review confirmed that there are no penetrations below the elevation of the direct vessel injection (DVI) nozzle that could lead to RCS water level dropping below the DVI nozzle. This provides assurance that the use of the CVS letdown line to achieve the reduction in RCS water inventory would not lead to the inadvertent uncovery of the core. Furthermore, the NRC staff also confirmed that the RCS water level under this condition would be approximately 4 ft above the top of active fuel elevation and that RCS coolant must be maintained under subcooled conditions consistent with the TS requirements for Mode 5 ( 200 °F). This provides assurance that the evaporative losses in the RCS would not be sufficient to uncover the core.

The low evaporation rate also assures that the boron concentration in the core would not increase to the precipitation limit.

The NRC staff evaluated the availability of controls necessary to prevent boron dilution in Mode 5 and 6. In Section 3 of the LAR, the applicant confirmed that consistent with the requirements in TS 3.4.8 for Mode 5 and in TS 3.9.2 for Mode 6, inadvertent boron dilution will be prevented by isolating all the sources of unborated water. Furthermore, TS 3.3.2, Reactor Trip System (RTS) Source Range Instrumentation, requires the source range nuclear instrumentation to be operable in Mode 5, which would detect any increase in the neutron flux due to potential boron dilution. In Mode 6, TS 3.9.3, Nuclear Instrumentation, requires two source range neutron flux

6 monitors to be operable, which would detect and give operators indication of any increase in the neutron flux due to potential boron dilution. These TSs provide assurance that there are adequate controls available to prevent boron dilution.

Therefore, it is reasonable that the IRWST injection and recirculation flow paths, the IRWST and ADS Stage 4 actuation instrumentation and controls, and the CVS automatic isolation function do not have any safety functions, and do not need to be operable prior to the core becoming irradiated. However, there are safety functions that require these functions once the fuel is irradiated. Accordingly, the proposed notes are structured so that they are only applicable prior to initial criticality. In Section 3 of the Enclosure to the LAR, the licensee states:

The term initial criticality is a commonly used term in the nuclear industry to refer to the time at which the reactor is first made critical. A reactor achieves criticality (and is said to be critical) when each fission event releases a sufficient number of neutrons to sustain an ongoing series of reactions. Initial criticality is an important milestone in the construction and commissioning of a nuclear power plant. Initial criticality is referred to repeatedly throughout the licensing basis documents, including the Combined License and Updated Final Safety Analysis Report (UFSAR), and its meaning is unambiguous, as there is a single defined point at which the reactor reaches criticality.

The staff reviewed VEGP Unit 3 UFSAR Section 14.2.7, Initial Fuel Loading and Initial Criticality, where the staff noted the UFSAR states that initial criticality follows initial core load. The staff agrees that the term initial criticality defines a specific time when the reactor is first made critical. Based on this, the staff finds that the proposed LCO notes provide a clear demarcation between the operating condition where IRWST does not need to be operable and the operating condition where IRWST must be operable. Once initial criticality is achieved, the notes will no longer be applicable and the affected LCO provisions will be required to be met.

Therefore, the LCOs continue to meet 10 CFR 50.36 because they continue to define the lowest functional capability or performance levels of equipment required for safe operation of the facility.

In addition, the licensee did not request, and this license amendment does not authorize, any changes to the approved design of Vogtle 3. While operating the plant in Modes 5 and 6 prior to initial criticality, no decay heat and other residual heat from the reactor core will be generated.

Accordingly, compliance with GDC 34 is not affected by this amendment. Similarly, the plant maintains its system (1) to automatically initiate systems to control reactivity to assure that SAFDLs are not exceeded as a result of AOOs and (2) to sense accident conditions and initiate operation of systems and components important to safety. The instrumentation and controls that will be disabled to make the repair are not required prior to initial criticality but must be operable once initial criticality is achieved, so compliance with GDC 20 is not affected by this amendment.

Therefore, the staff finds that the proposed changes are acceptable because:

1. The changes are consistent with applicable regulatory requirements. The requirements of 10 CFR 50.36 continue to be met. Compliance with GDC 20 and 34 is not affected.

2. The proposed changes are consistent with the accident analyses for Vogtle 3.

3. Operating the plant in accordance with the proposed TS changes provides reasonable assurance of adequate protection of public health and safety.

7 Error in Unit 3 Technical Specification page 3.3.88 The staff is addressing the errors described in Section 3.1 of this SE, in accordance with current policy and guidance. In SECY96238 (Ref. 8), the NRC staff informed the Commission of its intent to issue guidance to staff for determining what action is necessary to correct a typographical error associated with power reactor TSs. In a Staff Requirements Memorandum (Ref. 9), the Commission stated that it did not object to issuing the proposed guidance with certain changes identified by the Commission. The actual guidance was issued in an NRC memorandum (Ref. 10).

The typographical errors are editorial in nature and were introduced in Unit 3 Amendment 118 and inadvertently carried forward into Unit 3 Amendment 148 and the erroneous changes were not addressed in the notice to the public nor reviewed by the staff. In accordance with the previously discussed Commission-approved guidance and staff guidance in LIC101, Revision 6, License Amendment Review Procedures, (Ref. 11) the staff finds these errors to be typographical and is issuing the correction to Unit 3 TS page 3.3.88 with this amendment because the errors were not proper amendments in that they were not addressed in a notice to the public and not reviewed by the staff. Thus, the correction of these errors is not a change to the technical specifications as approved by the NRC. The name errors for Functions 18, 20, and 21 on Unit 3 TS page 3.3.88 are corrected in the change pages issued with this license amendment.

3.3

SUMMARY

In LAR 23004, SNC proposed to make changes that would affect TS LCOs for the Unit 3 IRWST injection valves and instrumentation not required during the drain down associated with the repair by making operability requirements not applicable prior to initial criticality of the fuel in the core. None of the above proposed changes represent changes to the design or construction of the plant. The NRC staff determined that the proposed TS changes satisfy the requirements of 10 CFR 50.36(c)(2)(i) because the LCOs continue to specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The staff also determined that compliance with GDC 34 is not affected because no decay heat will be generated prior to initial criticality. In addition, the amendment does not affect compliance with GDC 20 because the instrumentation systems that may be inoperable during the repair are not needed to prevent or mitigate an accident prior to initial criticality and the source range nuclear instrumentation is able to detect any increase in the neutron flux due to potential boron dilution.

In addition, facility operations in accordance with the LCOs can be conducted without endangering the health and safety of the public.

4.0 EXIGENT SITUATION The NRCs regulations in 10 CFR 50.91(a)(6) state that where the Commission finds that an exigent situation exists, in that the licensee and Commission must act quickly and time does not permit the Commission to publish in the Federal Register a notice allowing 30 days for prior public comment and also determines that the amendment involves no significant hazards considerations, it may issue a license amendment involving no significant hazards consideration with a local media notice and reasonable opportunity for the public to comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing an opportunity for a hearing. In addition, 50.91(a)(6) requires the licensee to explain the exigency and why it could not be avoided, and the staff to determine the appropriate notice and comment procedures upon consideration of the licensees basis for processing the proposed amendment

8 as an exigent amendment. This section documents the staffs consideration of the licensees basis.

As discussed in the licensees application dated January 31, 2023, the licensee requested that the proposed amendment be processed by the NRC on an exigent basis. Regarding the basis for the exigent situation, SNC stated the following, in part, in the LAR:

To maintain compliance with the current TS in order to complete the repair, SNC has developed three options: (i) a full core offload of the fuel assemblies to the spent fuel pool (thereby exiting MODE 6 and TS 3.5.8 Applicability), (ii) use of a freeze seal to isolate 3PXS-V123A from the borated RCS, or (iii) reduce RCS water level to below the DVI nozzle. Proceeding to a full core offload, and subsequent reloading of fuel assemblies, is estimated to add >16 days to the scheduled return to startup activities.

Implementation of a freeze seal and/or lowering RCS water level is preferred; however, the freeze seal provides only a single point isolation. In the event the currently planned use of a freeze seal to allow repair is not successful, without this amendment the plant would have to defuel to maintain compliance with current TS.

This repair activity is identified as an impediment for completing startup testing, which is the critical path activity for entry into Mode 2 (currently scheduled within a week of repairing 3PXS-V123A leakage) and the eventual operation of Unit 3. To minimize potential delays, SNC is requesting approval of this change, which will allow disabling IRWST injection and recirculation during the repair activities on Unit 3 IRWST injection isolation valve 3PXS-V123A leakage and allow the option of lowering reactor coolant system (RCS) level to below the direct vessel injection (DVI) nozzle elevation while maintaining compliance with the TS.

The NRC staff reviewed the licensees basis for processing the proposed amendment as an exigent amendment and has determined that an exigent situation exists consistent with the provisions in 10 CFR 50.91(a)(6). The NRC staff determined that: (1) the licensee has explained the basis for the exigency, (2) the licensee used its best efforts to make a timely application and could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91(a)(6). Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the exigent provisions of 10 CFR 50.91(a)(6).

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensees evaluation of the issue of no significant hazards consideration is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

9 Response: No The proposed changes do not adversely affect the operation of any structures, systems, or components (SSCs) associated with an accident initiator or initiating sequence of events. The proposed changes do not affect the design of the IRWST injection and recirculation flow paths, the reactor coolant system (RCS), or the associated PMS and DAS instrumentation.

The proposed amendment does not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility. The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the probability of an accident is increased.

The proposed amendment will not alter assumptions relative to the mitigation of an accident or transient event, as these assumptions are based upon irradiated fuel for the associated accident or transient. The proposed amendment does not increase the likelihood of the malfunction of an SSC or impact analyzed accidents.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment does not introduce any new or unanalyzed modes of operation. The proposed changes do not involve a physical alteration to the plant (i.e.,

no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis, as these assumptions are based upon irradiated fuel for the associated accident or transient.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment. The performance of these fission product barriers is not affected by the proposed amendment; therefore, the margins to the onsite and offsite radiological dose limits are not significantly reduced.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above evaluation, the staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the staff has made a final determination that no significant hazards

10 consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations in 10 CFR 50.91(b) the Georgia State official was notified of the proposed issuance of the amendment. In accordance with the Commissions regulations in 10 CFR 50.91(b), on February 5, 2023, the Commission consulted the State official. The State of Georgia had no comment.

7.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, Standards for Protection Against Radiation. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The NRC has made a final no significant hazards consideration determination, as stated above. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

8.0 CONCLUSION

The staff has concluded, based on the considerations discussed in Section 3.0 that there is reasonable assurance that: (1) the health and safety of the public will not be endangered by the proposed changes, (2) the changes are in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Therefore, the staff finds the changes proposed in this license amendment acceptable.

9.0 REFERENCES

1. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Unit 3, Exigent License Amendment Request: Technical Specification Exceptions for In-containment Refueling Water Storage Tank Operability Prior to Initial Criticality (LAR-23-004),

January 31, 2023 (ML23012A238).

2. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Unit 3, License Amendment No. 100, Addition of IRWST Lower Narrow Range Level Instrumentation, November 22, 2017 (package ML17284A066).

3. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Unit 3, License Amendment No. 118, Technical Specifications for Reactor Coolant System Vacuum Fill and Inspections, Tests, Analyses, and Acceptance Criteria for Containment Floodup, March 29, 2018 (package ML18075A094).

4. Not used.

11

5. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Unit 3, License Amendment No. 148, Engineered Safety Feature Safeguards Actuation Technical Specifications Applicability Changes, November 13, 2018 (ML18296A412)

6. Combined License NPF-91 for Vogtle Electric Generating Plant, Unit 3, Southern Nuclear Operating Company, April 10, 2014 (ML14100A106).

7. Vogtle Electric Generating Plant Units 3 and 4, Updated Final Safety Analysis Report, Revision 11, June 15, 2022 (ML22179A145).

8. NRC SECY-96-238, Proposed Guidance for Correction of Technical Specification Typographical Errors, November 19, 1996 (ML20134M324).

9. NRC Staff Requirements Memorandum - SECY-96-238, Proposed Guidance for Correction of Technical Specification Typographical Errors, December 17, 1996 (ML003754054).

10. NRC Memorandum License Amendment Corrections of Technical Specifications, January 16, 1997 (ML103260096).

11. LIC101, Revision 6, License Amendment Review Procedures, July 31, 2020 (ML19248C539).

NRC Staff Contributors Rob Elliott Pravin Sawant Joseph Ashcraft William Roggenbrodt