ML25136A381
| ML25136A381 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/17/2025 |
| From: | John Lamb NRC/NRR/DORL/LPL2-1 |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| EPID L-2025-LLA-0082 TS 3.7.6 | |
| Download: ML25136A381 (1) | |
Text
May 17, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company, Inc.
3535 Colonnade Parkway, Bin N-274-EC Birmingham, AL 35243
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT, UNIT 4 - ISSUANCE OF EMERGENCY AMENDMENT TO REVISE TECHNICAL SPECIFICATION 3.7.6, MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM (VES)
(EMERGENCY CIRCUMSTANCES) (EPID L-2025-LLA-0082)
Dear Ms. Coleman:
In response to your application dated May 16, 2025 (Agencywide Documents Access and Management System Accession No. ML25136A392), as supplemented by letter dated May 17, 2025 (ML25137A002), the U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 197 to Combined License (COL) No. NPF-92 for Vogtle Electric Generating Plant (Vogtle), Unit 4. The amendment revises Technical Specification (TS) 3.7.6, Main Control Room Emergency Habitability System (VES), to add a one-time allowance to provide time to repair VES bottled air system valve leakage.
The amendment is issued under emergency circumstances as described in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations due to the time critical nature of the amendment.
In this instance, an emergency exists because Vogtle, Unit 4, VES bottled air bank 2 outlet manual isolation valve, 4-VES-V025B, has been found with packing leakage. The repair of 4-VES-V025B requires complete isolation of bottled air from the VES air delivery header, rendering failure to meet TS 3.7.6 Required Action E.1. In that event, the repair options developed by Southern Nuclear Operating Company cannot be reasonably accomplished within the six hours allowed by Required Action F.1, and plant transition to Mode 3, Hot Shutdown, would be required within six hours, followed by Required Action F.2 requirement to be in Mode 5, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
A copy of the related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the final no significant hazards consideration determination. A Notice of Issuance addressing the final no significant hazards consideration determination and opportunity for a hearing associated with the emergency circumstances will be included in the Commissions monthly Federal Register notice.
J. Coleman If you have questions, please contact me at 301-415-3100 or John.Lamb@nrc.gov.
Sincerely,
/RA/
John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.: 52-026
Enclosures:
- 1. Amendment No. 197 to Vogtle, Unit 4, COL
- 2. Safety Evaluation cc: Listserv
SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 4 DOCKET NO.52-026 AMENDMENT TO FACILITY COMBINED LICENSE Amendment No. 197 License No. NPF-92
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Southern Nuclear Operating Company (SNC),
dated May 16, 2025, as supplemented by letter dated May 17, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I; B.
The facility will be constructed and will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.
2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.D(8) of Facility Combined Operating License No. NPF-92 is hereby amended to read as follows:
(8) Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 197, are hereby incorporated into this license.
3.
This license amendment is effective as of the date of its issuance and shall be implemented immediately.
FOR THE NUCLEAR REGULATORY COMMISSION:
Glenn E. Miller, Acting Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Page 7 of the facility Combined License and affected pages of Appendix C of the facility Combined License Date of Issuance: May 17, 2025 GLENN MILLER Digitally signed by GLENN MILLER Date: 2025.05.17 18:15:43 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 197 TO FACILITY COMBINED LICENSE NO. NPF-92 DOCKET NO.52-026 Replace the following pages of the facility Combined License No. NPF-92 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Combined License No. NPF-92 REMOVE INSERT 7
7 Appendix A to facility Combined License No. NPF-92 REMOVE INSERT 3.7.6-3 3.7.6-3
7 Amendment No. 197 (7)
Reporting Requirements (a)
Within 30 days of a change to the initial test program described in UFSAR Section 14, Initial Test Program, made in accordance with 10 CFR 50.59 or in accordance with 10 CFR Part 52, Appendix D, Section VIII, Processes for Changes and Departures, SNC shall report the change to the Director of NRO, or the Directors designee, in accordance with 10 CFR 50.59(d).
(b)
SNC shall report any violation of a requirement in Section 2.D.(3),
Section 2.D.(4), Section 2.D.(5), and Section 2.D.(6) of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made to the NRC Operations Center in accordance with 10 CFR 50.72, with written follow up in accordance with 10 CFR 50.73.
(8)
Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 197, are hereby incorporated into this license.
(9)
Technical Specifications The technical specifications in Appendix A to this license become effective upon a Commission finding that the acceptance criteria in this license (ITAAC) are met in accordance with 10 CFR 52.103(g) with the following exceptions:
(a)
Prior to initial criticality of the reactor core while operating in plant operational Mode 5 (Cold Shutdown) or Mode 6 (Refueling) the following TS are temporarily excluded from becoming effective:
TS 3.3.8, Engineered Safety Feature Actuation System (ESFAS)
Instrumentation, Table 3.3.8-1 o Function 14, RCS Wide Range Pressure - Low o Function 15, Core Makeup Tank (CMT) Level - Low 3 o Function 16, CMT Level - Low 6 o Function 18, IRWST Lower Narrow Range Level - Low 3 TS 3.3.9, Engineered Safety Feature Actuation System (ESFAS)
Manual Initiation, Table 3.3.9-1 o Function 1, Safeguards Actuation - Manual Initiation o Function 6, ADS Stages 1, 2 & 3 Actuation - Manual Initiation o Function 7, ADS Stage 4 Actuation - Manual Initiation o Function 8, Passive Containment Cooling Actuation - Manual Initiation o Function 9, Passive Residual Heat Removal Heat Exchanger Actuation - Manual Initiation
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 197 TO THE COMBINED LICENSE NO. NPF-92 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 4 DOCKET NO.52-026
1.0 INTRODUCTION
By letter dated May 16, 2025 (Agencywide Documents Access and Management System Accession No. ML25136A392), as supplemented by letter dated May 17, 2025 (ML25137A002),
Southern Nuclear Operating Company (SNC, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) amend Vogtle Electric Generating Plant (Vogtle), Unit 4, Combined License (COL) No. NPF-92.
The emergency license amendment request (LAR) proposed to revise Technical Specification (TS) 3.7.6, Main Control Room Emergency Habitability System (VES), under emergency circumstances, to add a one-time allowance to provide additional time to repair VES bottled air system valve leakage.
2.0 REGULATORY EVALUATION
2.0 REGULATORY EVALUATION
2.1 System Design and Operation As described in its submittal dated May 16, 2025, the licensee states, in part, that:
As described in the TS 3.7.6 Bases, there are 32 compressed air storage tanks arranged in four banks. Each bank is normally aligned to the VES air delivery header providing at least 81,893.5 scf of compressed air to support the minimum required 327,574 scf to support the safety design basis for automatic VES operation for 72-hours of post-accident operation. With the compressed air equivalent of one bank of tanks unavailable, verifying > 245,680 scf of compressed air ensures the equivalent of three banks remains available to supply air to the main control room envelope (MCRE) for 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> (75% of the required 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The MCRE ancillary fans are capable of maintaining the habitability of the MCRE after 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />.
2.2 Regulations and Guidance In accordance with paragraph C.6. of Section VIII of the Code of Federal Regulations (10 CFR)
Processes for Changes and Departures of Appendix D to Part 52 Design Certification Rule for the AP1000 Design, changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90. Pursuant to 10 CFR 50.90, whenever a COL holder desires to amend the license, application for an amendment must be filed with the Commission fully describing the changes desired and following, as far as applicable, the form prescribed for original applications. Per 10 CFR 52.79(a), an application for a COL must contain a final safety analysis report that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components of the facility as a whole.
Per 10 CFR 52.79(a)(11), the application for a COL shall include proposed TSs prepared in accordance with the requirements of 10 CFR 50.36. As required by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCO) that are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Further, per 10 CFR 50.36(c)(2)(i), 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. Per 10 CFR 50.36(c)(3), TSs will include surveillance requirements (SRs) that 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 LCO will be met. Per 10 CFR 52.97(c), a COL shall contain the terms and conditions, including TSs, as the Commission deems necessary and appropriate.
Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a)
(regarding, among other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other TSs, or the proposals) and those specifically for issuance of combined licenses in 10 CFR 52.97(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.
The regulation in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 19, Control room, states:
A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.
Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
3.0 TECHNICAL EVALUATION
3.1 Current TSs Current TS 3.7.6, Condition F, Required Action does not have a NOTE.
3.2 Proposed TSs The proposed added NOTE to TS 3.7.6, Condition F, Required Action would state:
NOTE-------------------
For one-time use on Unit 4, initiation of Required Actions F.1 and F.2 Completion Times may be delayed for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> from discovery of Condition F entry when Condition F entry is a result of VES air system leakage repair(s), provided VBS is providing ventilation to the main control room envelope. This provision expires 124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br /> following NRC approval.
3.3 Technical Evaluation of the Proposed Changes 3.3.1 Evaluation of Risk Insights The NRC staff reviewed the risk insights provided by the licensee regarding the proposed change to add a one-time allowance to provide time to repair VES bottled air system valve leakage.
SNC stated that its probabilistic risk assessment (PRA) model includes credit for the VES function to provide passive control room cooling capabilities, but that its PRA model does not
include the separate VES function to provide a passive air supply to the main control room because that is not a necessary function to prevent or mitigate core damage, and therefore, there is no quantifiable Core Damage Frequency (CDF) or Large Early Release Frequency (LERF) change associated with the proposed change.
The licensee further stated that based on the NRC Enforcement Manual, Appendix F, Notices of Enforcement Discretion (December 26, 2023, ADAMS Accession No. ML23362A014),
Section 1.4, that it believes the lessor impact to plant safety is provided by allowing sufficient time to complete valve leakage repairs at power rather than requiring the plant transient imposed by transitioning through multiple modes to accomplish a plant shutdown.
The NRC staff concludes that the available risk insights provided by SNC in Section 3 of the letter dated May 16, 2025, are acceptable for the purposes of supporting the deterministic evaluation. Because the VES function to provide a passive air supply to the main control room is not a necessary function to prevent or mitigate core damage, neither the licensee nor the NRC staff using its available models, could identify any quantifiable changes to CDF or LERF associated with the proposed change.
3.3.2 Radiological Dose Assessment The Vogtle, Unit 4, control room ventilation system exhibits defense-in-depth with the VBS ventilation system. While this is a non-safety-related system, the Vogtle, Units 3 and 4, updated final safety analysis report (UFSAR) table 15.6.5-3 provides the radiological consequence analysis of a Loss-of-Coolant Accident (LOCA) when relying on the VBS when VES is not available. This table provides results from both the safety-related system (VES) and the non-safety related backup system (VBS). This analysis demonstrates that the control room ventilation backup system, VBS, will be able to maintain control room dose below regulatory limits during the design basis LOCA.
With respect to the fission product inventory present in the source term used in the maximum credible accident (MHA) LOCA design basis dose analysis, Regulatory Guide 1.183 states:
The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full-power operation of the core with, as a minimum, currently licensed values for fuel enrichment and fuel burnup, and an assumed core power equal to the currently licensed rated thermal power times the approved core power measurement uncertainty factor (e.g., 1.02). These parameters should be examined to maximize fission product inventory.
The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach maximum values.
The licensee should not adjust the fission product inventory for events postulated to occur during power operations at less than full-rated power or those postulated to occur at the beginning of core life.
Vogtle, Unit 4, began commercial operation in April 2024. Since the core has not been subject to a single complete fuel cycle, none of the assemblies have experienced fuel burnup that would
allow the activity of dose-significant radionuclides to reach maximum values. Due to the time of commercial operation, none of the assemblies in the Vogtle, Unit 4, core are on their second or third burn cycle; therefore, the fission product inventory and, subsequently, the source term associated with the Vogtle, Unit 4, core, would be appreciably less than the values used in the current licensing basis analysis of record.
For this application, SNC is requesting a one-time change in allowed outage time to restore engineered systems to match the design and licensing basis of the plant. The licensee has demonstrated that the compensatory measures provided with VBS would allow the site to meet control room dose criteria. These back-up systems demonstrate defense-in-depth and redundancy. The VBS system is designed with redundant subsystems, and the VBS has a backup power supply in the event of a loss of offsite power. While in a situation where the fission product inventory available for release is appreciably less than the source term used in the current analysis of record, the proposed timeframe to take corrective action is commensurate with the significance of the nonconforming condition.
The NRC staff reviewed assumptions, inputs, and methods used by SNC to assess the radiological impacts of loss of VES concurrent with an MHA-LOCA, find that the values used in the calculations are consistent with their current licensing basis. Therefore, based on the above, the NRC staff finds the one-time extension of the completion time from 6 to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> proposed in the NOTE for Condition F is acceptable with respect to a radiological consequences analysis due to the available margin of the current Vogtle, Unit 4, source term compared to the design basis accident (DBA) analysis, and the compensatory measures to support availability of the VBS ventilation system.
The proposed TS 3.7.6 NOTE for Condition F allows for the one-time use of an extended completion time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to support VES air system leakage repair before Vogtle, Unit 4, is required to begin shutting down. Based on the evaluation above, the compensatory actions included in the proposed TS 3.7.6 NOTE provide reasonable assurance that the use of VBS when VES is unavailable will not significantly affect plant safety, because during the period the note is invoked, the main control room will continue to protect the control room personnel. The proposed 100-hour completion time is reasonable to complete repairs of the VES air system.
The NRC staff determined that the regulatory requirements of 10 CFR 50.36(c)(2) will continue to be met, because the modified TS will continue to describe the lowest functional capability or performance level of equipment required for safe operation of the facility and the remedial actions permitted by the TSs until the LCO can be met.
3.3.3 Valve Repair or Replacement The licensee has currently identified one valve (4-VES-V025B) that needs to be repaired or replaced as part of this license amendment activity. Four additional valves (4-VES-V025A, 4-VES-V024B, 4-VES-V027D, and 4-VES-V040D) may also require repair or replacement. Four of these five valves (three manual valves and one safety relief valve) are safety-related valves such that the repair/replacement activity will be performed in compliance with the quality assurance requirements of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. These four valves are within the scope of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section III, Rules for Construction of Nuclear Facility Components, such that the repair/replacement activities will be performed in accordance with the applicable requirements of the ASME BPV Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, as incorporated by reference in 10 CFR 50.55a, Codes and standards. The
applicable safety relief valve (4-VES-V040D) is within the scope of the Inservice Testing (IST)
Program required by the ASME Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code) as incorporated by reference in 10 CFR 50.55a, such that the repair/replacement activities will be performed in accordance with the applicable requirements of the ASME OM Code, Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants. The remaining manual valve being considered for repair or replacement is a non-safety-related valve such that the licensee will apply its applicable plant procedures for such activities. The licensee might identify additional valves needing repair or replacement during the implementation of this license amendment.
3.3.4 VBS Evaluation The licensee states in section 3 of enclosure to the LAR that VBS is designed to control the radiological habitability within the guidelines presented in Standard Review Plan 6.4, the main control room within the Control Room Habitability System, and NUREG 0696, Functional Criteria for Emergency Response Facilities, and within allowable GDC 19 limits during design basis accidents in both the main control room and control support area.
UFSAR section 6.4.4 states that in the event of an accident involving the release of radioactivity to the environment, VBS is expected to switch from the normal operating mode to the supplemental air filtration mode to protect the main control room personnel. Although VBS is not a safety-related system, it is expected to be available to provide the necessary protection for realistic events.
The licensee states following in section 3 of enclosure to the LAR: In the event of a loss of the normal ac power and offsite power, two standby diesel generators provide power to support the operation of VBS. Since VBS operates continuously during normal plant operation, its availability is continuously evident and maintaining its continued reliability is necessary for normal plant operations.
UFSAR section 1.9.4.2.3 states that in the event of external smoke or radiation release, VBS provides for a supplemental filtration mode of operation. In the unlikely event of a toxic chemical release, VES has the capability to be manually actuated by the operators. Further, a 6-hour supply of self-contained portable breathing equipment is stored inside the main control room pressure boundary. However, UFSAR section 6.4.4.2 states that toxic chemical habitability analysis results for offsite and site-specific onsite chemicals show that there are no toxic hazards to Units 3 and 4 control room personnel. As such, the licensee states in section 3 of enclosure to the LAR that the use of VES or self-contained portable breathing equipment is not postulated to be needed for toxic chemical releases. The TS operability requirement for VES does not include the necessity for VES to be operable for toxic gas event protection.
UFSAR section 9.4.1.2.3.1 states that if a high concentration of smoke is detected in the outside air intake, an alarm is initiated in the main control room and VBS is manually realigned to the recirculation mode by closing the outside air and toilet exhaust duct isolation valves. The exhaust fans for the main control room toilets and the toilet serving the control support area are tripped upon closure of the isolation valves. As such, the licensee states in section 3 of enclosure to the LAR that the TS operability requirement for VES does not include the necessity for VES to be operable for smoke event protection.
As stated in section 3.2 of this report, the licensees proposed NOTE to TS 3.7.6, Condition F, Required Action allowing for a one-time use of an extended completion time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to support VES air system leakage repair is conditional on VBS providing ventilation to the MCRE.
Based on the review of information presented in the LAR, TS, and UFSAR, the NRC staff determines that the licensees proposed request allowing for a one-time use of an extended completion time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to TS 3.7.6, Condition F, Required Action to support VES air system leakage repair is acceptable, provided VBS is available to provide ventilation to the MCRE.
3.3.5 Evaluation of TS Change The proposed TS 3.7.6 NOTE for Condition F allows for the one-time use of an extended completion time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to support VES air system leakage repair before Vogtle, Unit 4, is required to begin shutting down, provided VBS is available to provide ventilation to the MCRE.
The compensatory actions included in the proposed TS 3.7.6 NOTE provide reasonable assurance that the use of VBS when VES is unavailable will not significantly affect plant safety because during the period the NOTE is invoked, the main control room will continue to protect the control room personnel. The proposed 100-hour completion time is reasonable to complete repairs of the VES air system. The NRC staff determined that the regulatory requirements of 10 CFR 50.36(c)(2) will continue to be met, because the modified TS will continue to describe the lowest functional capability or performance level of equipment required for safe operation of the facility and the remedial actions permitted by the TSs until the LCO can be met.
4.0 EMERGENCY CIRCUMSTANCE The NRCs regulations in 10 CFR 50.91(a)(5) state where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, the NRC may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.
As discussed in Section 2.3 of SNCs submittal dated May 16, 2025, SNC requested that the proposed LAR be reviewed by the NRC on an emergency basis, because Vogtle, Unit 4, VES bottled air Bank 2 outlet manual isolation valve, 4-VES-V025B, has been found with packing leakage. The repair of 4-VES-V025B requires complete isolation of bottled air from the VES air delivery header, rendering failure to meet TS 3.7.6 Required Action E.1. In that event, the repair options developed by SNC cannot be reasonably accomplished within the six hours allowed by Required Action F.1, and plant transition to Mode 3, Hot Shutdown, would be required within six hours, followed by Required Action F.2 requirement to be in Mode 5, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
NRC Staff Conclusion regarding Emergency Circumstance The NRC staff reviewed the licensees basis for processing the proposed LAR as an emergency amendment, as discussed above, and has determined that an emergency situation exists consistent with the provisions in 10 CFR 50.91(a)(5). Furthermore, the NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of
10 CFR 50.91(a)(5). Based on these findings, and the determination that the LAR involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the LAR using the emergency provisions of 10 CFR 50.91(a)(5).
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not affect the previously evaluated accident probability because the Updated Final Safety Analysis Report (UFSAR)
Chapter 15 initiating events for analyzed accidents do not change. The main control room emergency habitability system (VES) is not considered an accident initiator. The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated or maintained.
Therefore, the proposed change does not result in any increase in probability of an analyzed accident occurring.
The proposed change does not impact the nuclear island nonradioactive ventilation system (VBS) ability to protect control room personnel from radiological consequences. The proposed change provides that the main control room will continue to protect the control room personnel in meeting General Design Criterion (GDC) 19 dose limits. Thus, the consequences of the accidents previously evaluated are not adversely affected.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?
Response: No.
The proposed change does not involve the installation of any new or different type of equipment or a change to the methods governing normal plant operation. The proposed changes continue to provide the functional capability for previously evaluated accidents and anticipated operational occurrences. The proposed
change does not adversely impact the function of any related systems, and thus, the changes do not introduce a new failure mode, malfunction, or sequence of events that could adversely affect safety or safety-related equipment Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. The proposed amendment does not alter the design or configuration of the VES or VBS. The proposed change does not change the function of the related systems. The proposed change was evaluated and demonstrated that the control room habitability during analyzed events remains.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
For the reasons noted above, there is no significant reduction in a margin of safety. Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards 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 Commissions regulations, the Georgia State official was notified of the proposed issuance of the amendments on May 16, 2025. There were no comments received from the State of Georgia official.
7.0 ENVIRONMENTAL CONSIDERATION
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.
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 Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: J. Wilson, NRR H. Wagage, NRR D. Coy, NRR S. Meighan, NRR J. Robinson, NRR T. Scarbrough, NRR N. Hansing, NRR Date: May 17, 2025
ML25136A381 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/SCPB/BC NRR/DSS/STSB/BC NAME JLamb KZeleznock MValentin SMehta DATE 05/16/2025 05/17/2025 05/17/2025 05/17/2025 OFFICE NRR/DEX/EMIB/BC NRR/DRA/APLB/BC NRR/DRA/ARCB/BC OGC - NLO NAME TScarbrough for SBailey JRobinson for EDavidson KHsueh SVrahoretis DATE 05/17/2025 05/17/2025 05/17/2025 05/17/2025 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME EMiller for MMarkley JLamb DATE 05/17/2025 05/17/2025