ML25342A206
| ML25342A206 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 12/30/2025 |
| From: | Thomas Scarbrough, Hanry Wagage Plant Licensing Branch II |
| To: | Carr E South Carolina Electric & Gas Co |
| Miller, GE | |
| References | |
| EPID L-2025-LLA-0054 | |
| Download: ML25342A206 (0) | |
Text
December 30, 2025 Mr. Eric S. Carr Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 ISSUANCE OF AMENDMENT NO. 228 TO MODIFY TECHNICAL SPECIFICATION 6.8.4.g TO MODIFY THE CONTAINMENT LEAKAGE RATE TESTING PROGRAM (EPID L-2025-LLA-0054)
Dear Mr. Carr:
The U.S. Nuclear Regulatory Commission (NRC, or the Commission) has issued the enclosed Amendment No. 228 to Subsequent Renewed Facility Operating License No. NPF-12 for the Virgil C. Summer Nuclear Station, Unit 1. The amendment revises the Technical Specifications (TSs) in response to your application dated March 20, 2025), as supplemented by letter dated September 5, 2025.
The amendment revises TS 6.8.4.g to modify the Containment Leakage Rate Testing Program.
The change replaces the reference to Nuclear Energy Institute (NEI) Technical Report 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option 10 CFR 50, Appendix J, October 2008, with NEI Technical Report 94-01, Revision 3-A, July 2012, and conditions and limitations specified in NEI 94-01, Revision 2-A.
E. Carr A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
G. Edward Miller, Project Manager Plant Licensing Branch II-I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-395
Enclosures:
- 1. Amendment No. 228 to NPF-12
- 2. Safety Evaluation cc: Listserv DOMINION ENERGY SOUTH CAROLINA, INC.
SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 228 Renewed License No. NPF-12
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Virgil C. Summer Nuclear Station, Unit No. 1 (the facility), Subsequent Renewed Facility Operating License No. NPF-12, filed by the Dominion Energy South Carolina, Inc. (the licensee), dated March 20, 2025, as supplemented by letter dated September 5, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations 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 public health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations as 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 Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by a page change to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Subsequent Renewed Facility Operating License No. NPF-12 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the subsequent renewed license.
Dominion Energy South Carolina, Inc. 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 within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: December 30, 2025 SHAWN WILLIAMS Digitally signed by SHAWN WILLIAMS Date: 2025.12.30 13:26:51 -05'00'
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 ATTACHMENT TO LICENSE AMENDMENT NO. 228 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the subsequent renewed facility operating license with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Page Insert Page License License Page 3 Page 3 Technical Specifications Technical Specifications 6-12b 6-12b
Renewed Facility Operating License No. NPF-12 Amendment No. 228 (3)
DESC, 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 amounts required for reactor operation, as described in the Final Safety Analysis Report, as amended through Amendment No. 33; (4)
DESC, pursuant to the Act and 10 CFR Part 30, 40 and 70 to receive, possess and use at any time byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
DESC, 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 of components; and (6)
DESC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as m[a]y be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain, and is subject to, the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I 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 DESC is authorized to operate the facility at reactor core power levels not in excess of 2900 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to this renewed license.
The preoccupation tests, startup tests and other items identified in to this renewed license shall be completed as specified. is hereby incorporated into this renewed license.
(2)
Technical Specifications and Environmental Protection Plant The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. Dominion Energy South Carolina, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
SUMMER - UNIT 1 6-12b Amendment No.104, 117, 130, 135, 189, 194, 210, 228 ADMINISTRATIVE CONTROLS f.
Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measures of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
1)
Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM; 2)
A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of the census; and 3)
Participation in an Inter-laboratory Comparison Program to ensure that independent checks on the precision and accuracy of measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
g.
Containment Leakage Rate Testing Program A program shall be established to implement leakage rate testing of the containment system as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01 Industry Guideline for Implementing Performance-Based Option of 10CFR50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 46.0 psig.
The maximum allowable containment leakage rate, La, at Pa, is 0.20 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Leakage rate acceptance criteria are:
1)
Containment overall leakage rate acceptance criterion is 1.0 La.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests;
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 228 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. NPF-12 DOMINION ENERGY SOUTH CAROLINA, INC.
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 DOCKET NO. 50-395
1.0 INTRODUCTION
By application dated March 20, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML25079A198), as supplemented by letter dated September 5, 2025 (ADAMS Accession No. ML25248A244), Dominion Energy South Carolina, Inc. (DESC, the licensee) requested changes to the Technical Specifications (TSs) for the Subsequently Renewed Facility Operating License No. NPF-12 for Virgil C. Summer Nuclear Station (VCSNS), Unit 1. The proposed amendment would revise TS 6.8.4.g to modify the Containment Leakage Rate Testing Program. The proposed change replaces the reference to Nuclear Energy Institute (NEI) Technical Report 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option 10CFR50, Appendix J, October 2008, with NEI Technical Report 94-01, Revision 3-A, July 2012, and conditions and limitations specified in NEI 94-01, Revision 2-A.
The supplemental letter dated September 5, 2025, provided additional information that clarified the application did not expand the scope of the application as originally noticed, and did not change the U. S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 14, 2025 (90 FR 20516).
1.1 Description of Changes The proposed change revises VCSNS TS 6.8.4.g, by replacing the reference to NEI 94-01, Revision 2-A, with a reference to NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the documents used by the licensee to implement the performance-based leakage testing program in accordance with Title 10, Energy, of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance-Based Requirements.
The proposed change, as stated in section 1 of attachment 1 to the license amendment request (LAR), revises VCSNS TS 6.8.4.g, Containment Leakage Rate Testing Program to reflect the following:
Adopts an extension of the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.
Adopts a more conservative allowable test interval extension of nine months, for Type A, Type B, and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.
The proposed change would not impact the Type A integrated leakage rate test (ILRT) program test interval, which was previously extended to a permanent 15-year interval under Amendment No. 194,1 and maintains the conditions and limitations specified in NEI 94-01, Revision 2-A, required by that amendment.
2.0 REGULATORY EVALUATION
2.1
System Description
In section 3.2 of attachment 1 to the LAR, the licensee states in part:
VCSNS Containment is provided by the Reactor Building, a reinforced concrete structure designed by Gilbert Associates, Inc. The Reactor Building is a post tensioned, reinforced concrete structure with an integral steel liner. The Reactor Building consists of a cylindrical wall, a shallow dome roof, and a foundation mat with a depressed incore instrumentation pit under the reactor vessel. The foundation mat and cylindrical wall are reinforced with conventional mild steel reinforcing. The cylindrical wall is prestressed in the vertical and horizontal directions by a post-tensioning system. The shallow dome roof is prestressed by a three-way post-tensioning system. The inside surface of the Reactor Building is lined with a carbon steel liner to ensure a high degree of leak tightness under operating and accident conditions.
2.2 Regulatory Requirements The NRC regulations in 10 CFR 50.54(o) require that the primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in 10 CFR Part 50, Appendix J. This appendix specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components, which penetrate the containment. In addition, the appendix discusses leakage rate acceptance criteria, test methodology, frequency of testing, and reporting requirements for each type of test.
Appendix J to 10 CFR Part 50 includes two options: Option APrescriptive Requirements and Option BPerformance-Based Requirements, either of which can be chosen to meet the requirements of the appendix. The testing requirements in the appendix ensure that (a) leakage 1 Amendment 194 was issued on February 5, 2014 (ADAMS Accession No. ML13326A204).
through containments or systems and components penetrating containments does not exceed the allowable leakage rates specified in the TS and (b) the integrity of the containment structure is maintained throughout the service life of the containment.
VCSNS Unit 1 adopted Option B in TS 6.8.4.g, Containment Leakage Rate Testing Program.
The adoption of the Option B performance-based containment leakage rate testing for Type A, B, and C testing does not alter the basic method by which Appendix J leakage rate testing is performed. However, it does alter the frequency at which Type A, B, and C containment leakage tests must be performed. Under the performance-based option of 10 CFR Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed the as-found leakage history to determine the frequency for leakage testing, which provides assurance that leakage limits will be maintained.
In 10 CFR 50.36, Technical specifications, the NRC established its regulatory requirements related to TSs. Pursuant to 10 CFR 50.36(c)(1)-(5), TSs 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; (3) surveillance requirements; (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plants TSs. The regulation in 10 CFR 50.36(c)(5) requires TS to include items in the Administrative controls category, stating in part:
Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
ASME Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code), Subsection ISTC, Inservice Testing of Valves in Water-Cooled Reactor Nuclear Power Plants, paragraph ISTC-3620, Containment Isolation Valves, requires containment isolation valves (CIVs) with leakage rate requirements based on 10 CFR Part 50, Appendix J, to be tested in accordance with the Owners Appendix J program. For VCSNS, this ensures that CIVs subject to the Type C local leakage rate test (LLRT) program are governed by both the ASME OM Code and 10 CFR Part 50, Appendix J.
2.3 Regulatory Guidance NEI 94-01, Revision 0 (ML11327A025), provides methods for complying with the provisions of 10 CFR Part 50, Appendix J, Option B, and includes provisions that address the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests. VCSNS Unit 1 adopted Option B of 10 CFR Part 50, Appendix J, for integrated (Type A) and local (Types B and C) leakage rate testing with Amendment No. 135 (ML012270146) and is part of the current VCSNS Unit 1 licensing basis.
NEI 94-01, Revision 2-A, incorporates the regulatory positions stated in Regulatory Guide (RG) 1.163 and delineates a performance-based approach for determining Types A, B, and C containment leakage rate testing frequencies. It also includes provisions for extending Type A ILRT intervals to up to 15 years. This approach uses industry performance, plant-specific data, and risk insights in determining the appropriate testing frequency, and discusses the performance factors that licensees must consider in determining test intervals. In a letter dated June 25, 2008 (ML081140105), the NRC published a safety evaluation with limitations and conditions for NEI 94-01, Revision 2, and Electric Power Research Institute (EPRI) Report No. 1009325, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A (ADAMS Package No. ML072970204).
NEI 94-01, Revision 3-A, provides guidance for extending Type C LLRT intervals beyond 60 months. The NRC published a safety evaluation (SE) with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012 (ML121030286). In the SE, the NRC concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, in July 2012.
RG 1.163, Revision 1, endorses the guidance in NEI 94-01, Revision 3-A for implementing Option B of Appendix J to 10 CFR Part 50, subject to the regulatory positions listed in section C of this RG. This guidance includes (1) extending Type A test intervals up to 15 years and (2) extending Type C test intervals up to 75 months. These regulatory positions are the same as those listed in the NRC safety evaluation for NEI 94-01, Revision 2-A, which are missing in the NEI 94-01, Revision 3-A.
3.0 TECHNICAL EVALUATION
Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The staff evaluated the request to determine whether the proposed changes are consistent with the regulations, licensing basis, and guidance discussed in this safety evaluation. The NRC staff reviewed the proposed changes to TS to determine whether they meet the requirements of Appendix J to 10 CFR Part 50, 10 CFR 50.55a, and 10 CFR 50.36, and provide reasonable assurance that operation with the proposed changes will not endanger the health and safety of the public.
3.1 Technical Analysis 3.1.1. Primary Containment Leakage Rate Testing Program - Type B and Type C Testing Program In the LAR as updated in its supplement dated September 5, 2025, the licensee describes the Primary Containment Leakage Rate Testing Program for electrical penetrations, airlocks, hatches, flanges, and containment isolation valves in accordance with 10 CFR Part 50, Appendix J, Option B, and NEI 94-01. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. The Type B and Type C testing program provides a means to protect the health and safety of plant personnel and the public by maintaining leakage from these components below appropriate limits. In accordance with TS 6.8.4.g, the allowable maximum containment leakage rate (La) pathway total Types B and C leakage is 0.6 La that equals 176,823 standard cubic centimeters per minute (sccm).
In the LAR, the licensee asserts that Type B and Type C tests can identify most potential containment leakage paths such that Type B and Type C testing provides a high degree of assurance that containment integrity is maintained. According to the licensee, the Type B and Type C test results from 2014 through 2024 for VCSNS have shown substantial margin between the actual As-Found and As-Left outage summations and the regulatory requirements as follows:
The As-Found minimum pathway leak rate average for VCSNS Unit 1 shows an average of approximately 20% of 0.6 La with a high of 24% of 0.6 La.
The As-Left maximum pathway leak rate average for VCSNS Unit 1 shows an average of approximately 33% of 0.6 La with a high of 44% of 0.6 La.
As described in the following paragraphs of this SE section, the NRC staff confirmed the leak rates from the results provided in Table 2-1, Types B and C LLRT Combined As-Found/As-Left Trend Summary, in Attachment 2, Local Leak Rate Test Data Trend Summary and Program Implementation Review, to the LAR. Based on its review, the NRC staff finds that the licensee is effectively implementing the VCSNS Type B and Type C leakage rate test program, as required by Option B of 10 CFR Part 50, Appendix J. Therefore, extending the CIV leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR Part 50, Appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with the guidance in NEI 94-01, Revision 3-A, is acceptable. The NRC staff also finds that the licensee is adopting an allowable test interval extension of nine months, for Type A, Type B, and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.
3.1.2 Type B and Type C Local Leak Rate Program Implementation Review Table 2-1 in Attachment 2 to the LAR provides LLRT data trend summaries for VCSNS since 2014, inclusive of the 2018 ILRT. This table shows that there has been no As-Found failure that resulted in exceeding the TS 6.8.4.g limit of 0.6 La and demonstrates a history of successful tests. The As-Found minimum pathway summations support the adequacy of maintenance of Type B and Type C tested components, while the As-Left maximum pathway summations support the adequacy of the Containment Leakage Rate Testing Program. Table 2-2, Types B and C LLRT Program Implementation Review, in Attachment 2 to the LAR identifies the components that have not demonstrated acceptable performance during the previous three outages for VCSNS.
In its LAR review, the NRC staff performed the following to assess the technical adequacy of the proposed extension of the leak test intervals:
Examined the licensees 2014 to 2024 LLRT performance data provided in the LAR to determine whether adjusted leakage totals remained within the 10 CFR Part 50, Appendix J acceptance criteria.
Evaluated the licensees application of a 1.25 adjustment factor to As-Left test results for valves on intervals greater than 60 months, as described in the LAR supplement dated September 5, 2025 (ML25248A244), to verify that it provides a conservative bounding estimate of potential leakage.
Reviewed trend analyses provided by the licensee in the LAR supplement to confirm that performance monitoring demonstrates stable or improving leakage performance and that corrective actions are taken as needed.
Assessed aging and structural integrity considerations, including the licensees reliance on containment aging management programs discussed in the LAR supplement.
Verified that the licensees visual inspection commitments meet NEI 94-01, Revision 2-A, Condition 2, and are consistent with the ASME BPV Code,Section XI, requirements, as discussed in the LAR supplement.
The containment inservice inspection program implemented in accordance with ASME BPV Code,Section XI, Subsections IWE and IWL, as incorporated by reference in 10 CFR 50.55a, provides assurance of the structural integrity of containment pressure boundaries, including penetrations that house CIVs, and complements the Appendix J leakage rate testing program.
In its LAR supplement dated September 5, 2025, the licensee explained that aging effects for CIVs on extended test intervals are managed under the existing containment aging management program, as described in the VCSNS Subsequent License Renewal Application (SLRA, ML23233A172) and evaluated in the applicable NRC Safety Evaluation Report (ADAMS Package No. ML25059A003). Consistent with NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, these programs include seal and gasket replacement, corrosion prevention through chemistry controls, fatigue monitoring, and integration with ASME BPV Code,Section XI, Subsections IWE and IWL, inspections. Based on its review, the NRC staff determined that these planned actions provide adequate assurance of aging management for CIVs subject to the extended Type C LLRT interval and satisfy Condition 3 of NEI 94-01, Revision 2-A.
During its review, the NRC staff requested the licensee to confirm that at least three general visual inspections of the containment are performed during each 15-year Type A ILRT interval, consistent with NEI 94-01, Revision 2-A, Condition 2. In its LAR supplement dated September 5, 2025, the licensee confirmed this statement. Further, the NRC staff verified that the inspection program ensures that no more than 5 years elapse between inspections.
The NRC staff reviewed the 10 years of LLRT performance data provided in Attachment 2, Table 2-1, of the LAR. The NRC staff determined that the data show acceptable As-Left leakage values below administrative limits and within the combined 0.6 La threshold. The data demonstrate that even with adjustment, the highest observed values remain under 45% of the 0.6 La limit such as to provide substantial margin.
To account for potential leakage understatement from valves tested beyond 60 months, the licensee applied a 1.25 adjustment factor. This approach is consistent with NRC Safety Evaluation Condition 2 on NEI 94-01, Revision 3-A, and has been accepted by the NRC for similar LARs submitted by licensees of other nuclear power plants. The NRC staff reviewed this methodology and determined that it provides a conservative bounding estimate of leakage potential. In its LAR supplement dated September 5, 2025, the licensee described its plans for reporting post-outage leakage totals, including application of the 1.25 adjustment factor, and implementing corrective actions if margins to the 0.6 La limit are reduced. The licensees response further clarified that valves exceeding the acceptance criteria would have their test intervals reset in accordance with NEI 94-01, Revision 3-A. The licensee also stated that approximately two-thirds of Type C tested valves at VCSNS are currently on 60-month intervals and, based on satisfactory performance, may be considered for extension to 75 months on a case-by-case basis.
Based on the review of the LAR and its supplement, the NRC staff finds that the licensees implementation of the VCSNS Type B and Type C Local Leak Rate Program is adequate to provide reasonable assurance that containment integrity will be maintained under the extended interval leak testing program.
3.1.3 Type B and Type C Component Performance Summary In the LAR, the licensee stated that the percentage of the total number of VCSNS Unit 1 Type B tested components (70 total nozzles/airlocks associated with 61 penetrations) that are on 120-month extended performance-based test intervals is approximately 81% (57 components associated with 48 penetrations). The remaining Type B penetrations that are not on an extended test frequency are:
Used during refueling outages (RFOs) and therefore must be As-Left tested each RFO subsequent to use (8 components, approximately 11%),
As-Left tested each RFO to support FLEX requirements (2 components, approximately 3%), or Limited to a test frequency of 30 months per the Primary Containment Leakage Rate Testing Program (3 components, approximately 4%).
The percentage of the total number of VCSNS Unit 1 Type C tested components (100 total valves associated with 49 penetrations) that are on 60-month extended performance-based test intervals is approximately 66% (66 valves associated with 32 penetrations). The remaining Type C penetrations not on a 60-month extended test frequency are:
On a 30-month frequency (resulting in a test every RFO) following valve replacement or major maintenance to re-establish their performance history of two satisfactory consecutive As-Found tests (9 valves associated with 4 penetrations identified in, Table 2-2, approximately 9%),
Used or removed during RFOs to support FLEX or outage requirements (10 valves associated with 7 penetrations, approximately 10%), or Not on an extended test frequency of 60 months due to limitations by another program (e.g., Inservice Testing Program), even though they have met the performance requirements of two satisfactory consecutive As-Found test after the last Type C test (15 valves associated with 6 penetrations, approximately 15%).
Therefore, the licensee asserted that the current Type B and Type C performance supports an allowance to extend the leak test interval up to 75 months.
In its LAR supplement dated September 5, 2025, the licensee provided a graphical and tabular trend evaluation of Type C LLRT results over the past 10 years, including identification of recurring valve failures, corrective actions, and justification for continued eligibility for extended leak test intervals. The As-Left leakage totals across the last six outages ranged from approximately 0.125 La to 0.266 La with no adverse trends observed. The LAR supplemental information demonstrated that, even with application of the 1.25 adjustment factor, the leakage results remain well within allowable limits, and corrective actions have been effective in addressing isolated degradations. As shown in the LAR supplement, each instance of elevated leakage was promptly corrected, valves were returned to service with satisfactory results, and intervals were reset as required by NEI 94-01, Revision 3-A. The NRC staff has determined that the licensees trending program demonstrates stable leakage performance and that corrective actions have been effective.
Based on its review, the NRC staff finds that the Type B and Type C component performance summary is acceptable with the provisions in the LAR and its supplement to:
Continue applying the 1.25 adjustment factor to As-Left leakage values for components tested on extended intervals; Report post-outage combined Type B and C leakage totals, including adjusted values, against the 0.6 La regulatory limit; Initiate corrective actions if performance margins deteriorate, including resetting valve test intervals where needed; and Maintain at least three general containment inspections within each 15-year ILRT interval, consistent with NEI 94-01, Revision 2-A, Condition 2, and ASME BPV Code,Section XI, Subsections IWE and IWL.
3.2 Limitations and Conditions in NRC Safety Evaluation of NEI 94-01, Revision 3-A In the June 8, 2012 (ML121030286), SE for NEI 94-01, Revision 3, the NRC staff concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions. The SE was incorporated into Revision 3, and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012 (ML12221A202).
3.2.1 NEI 94-01, Revision 3-A, Condition 1 Condition 1 identifies three issues that are required to be addressed:
Issue 1 - The allowance of an extended interval for Type C LLRTs of 75 months requires that a licensees post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.
Issue 2 - A corrective action plan is to be developed to restore the margin to an acceptable level.
Issue 3 - Use of the allowed 9-month extension for eligible Type C valves is only allowed for non-routine emergent conditions, but not for valves categorically restricted and other excepted valves.
The licensees response to Condition 1, Issue 1, stated:
The post-outage report shall include the margin between the Type B and Type C
[as-left minimum pathway leakage rate (MNPLR)] summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.6 La.
The licensees response to Condition 1, Issue 2, stated:
VCSNS will assess and monitor margin between the Type B and C MNPLR summation value, as adjusted to include the estimate of applicable Type C leakage understatement, included in the post-outage report and its regulatory limit of 0.6 La. This will include corrective actions to restore margin to an acceptable level, if required. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action, as deemed appropriate, that best focuses on the prevention of future component leakage performance issues to maintain an acceptable level of margin.
The licensees response to Condition 1, Issue 3, stated:
VCSNS will apply the 9-month allowable interval extension period only to eligible Type C components for non-routine emergent conditions. Such occurrences will be documented in the record of tests.
The NRC staff reviewed the licensees responses and found that each of the three issues has been satisfactorily addressed without deviation, and therefore, Condition 1 of the NEI 94-01, Revision 3-A, SE has been satisfactorily addressed.
3.2.2 NEI 94-01, Revision 3-A, Condition 2 Condition 2 identifies two issues that are required to be addressed:
Issue 1 - Extending the Type C LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative, provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, section 12.1.
Issue 2 - When routinely scheduling any LLRT valve interval beyond 60-months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of penetrations.
The licensees response to Condition 2, Issue 1, stated:
The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94 01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, VCSNS will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As Left leak rate, which will increase the As Left leakage total for each Type C component currently on greater than a 60 month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being carried forward and will be included whenever the total leakage summation is required to be updated (either while online or following an outage).
The licensees response to Condition 2, Issue 2, states, in part:
The potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75 month extended test interval will be summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components. The margin between the MNPLR and the regulatory limit of 0.6 La will be assessed and monitored and included as part of the post-outage report.
The NRC staff reviewed the licensees responses and found that each of the three issues has been satisfactorily addressed without deviation, and therefore, Condition 2 of the NEI 94-01, Revision 3-A, SE has been satisfactorily addressed.
3.3 Technical Evaluation Conclusion
As described above, the NRC staff finds that the licensees LLRT program shows acceptable performance supporting the conclusion that the primary containment structural and leak-tight integrity is adequately managed and will continue to be periodically monitored and managed effectively with the proposed changes. The NRC staff finds that the licensee has addressed the NRC limitations and conditions identified in NEI 94-01, Revision 3-A.
Therefore, the NRC staff concludes that the proposed changes to VCSNS TS 6.8.4.g regarding the containment leakage rate testing program are acceptable and continue to meet 10 CFR 50.36(c)(5).
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the NRC staff notified the South Carolina State official of the proposed issuance of the amendment on September 16, 2025. On September 17, 2025, the State official confirmed the State of South Carolina had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component 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, and there has been no public comment on such finding published in the Federal Register on May 14, 2025 (90 FR 20516).
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 amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that public health and safety will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to public health and safety.
Principal Contributors: H. Wagage, NRR T. Scarbrough, NRR Date of issuance: December 30, 2025
ML25342A206 NRR-058 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NRR/DSS/SCPB/BC NAME GEMiller KZeleznock SMehta MValentin DATE 12/11/2025 12/11/2025 12/12/2025 12/16/2025 OFFICE NRR/DEX/EMIB/BC NRR/DEX/ESEB NRR/DORL/LPL2-1/BC NAME SBailey (TScarbrough for) ITseng MMarkley (SWilliams for)
DATE 12/17/2025 12/23/2025 12/30/2025