ML26027A354

Issuance of Amendments Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies

ML26027A354
Person / Time
Site:	Oconee   (DPR-038, DPR-047, DPR-055)
Issue date:	02/27/2026
From:	Shawn Williams Plant Licensing Branch II
To:	Snider S Duke Energy Carolinas
Williams S
References
EPID L-2025-LLA-0136
Download: ML26027A354 (0)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 27, 2026 Mr. Steven M. Snider Vice President, Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 - ISSUANCE OF AMENDMENTS RE: REVISE TECHNICAL SPECIFICATION 5.5.2, CONTAINMENT LEAKAGE RATE TESTING PROGRAM, FOR PERMANENT EXTENSION OF TYPE A AND TYPE C LEAK RATE TEST FREQUENCIES (EPID L-2025-LLA-0136)

Dear Mr. Snider:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 433, 435, and 434 to Subsequent Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station, Units 1, 2, and 3, respectively.

The amendments are in response to the application from Duke Energy Carolinas, LLC, dated August 19, 2025.

The amendments revise Technical Specification (TS) 5.5.2, Containment Leakage Rate Testing Program, to allow the following:

Increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years, Adopt an extension of the containment isolation valve (CIV) leakage rate testing frequency for Type C leakage rate testing of selected components, Adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2020, Containment System Leakage Testing Requirements, and Adopt 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.

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

If you have any questions, please email at shawn.williams@nrc.gov.

Sincerely,

/RA/

Shawn Williams, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287

Enclosures:

1. Amendment No. 433 to DPR-38

2. Amendment No. 435 to DPR-47

3. Amendment No. 434 to DPR-55

4. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 433 Subsequent Renewed License No. DPR-38

1.

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

A.

The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Subsequent Renewed Facility Operating License No. DPR-38, filed by Duke Energy Carolinas, LLC (the licensee), dated August 19, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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 page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 433, are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 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 No. DPR-38 and the Technical Specifications Date of Issuance: February 27, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.02.27 09:25:27 -05'00'

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 435 Subsequent Renewed License No. DPR-47

1.

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

A.

The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Subsequent Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated August 19, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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 page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 435, are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 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 No. DPR-47 and the Technical Specifications Date of Issuance: February 27, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.02.27 09:26:31 -05'00'

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 434 Subsequent Renewed License No. DPR-55

1.

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

A.

The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Subsequent Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), August 19, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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 page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 434, are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 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 No. DPR-55 and the Technical Specifications Date of Issuance: February 27, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.02.27 09:27:16 -05'00'

Attachment ATTACHMENT TO LICENSE AMENDMENT NOS. 433, 435, AND 434 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-38, DPR-47, AND DPR-55 DOCKET NOS. 50-269, 50-270, AND 50-287 Replace the following pages of the Subsequent Renewed Operating Licenses and the Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages Operating Licenses Operating Licenses License No. DPR-38, page 3 License No. DPR-38, page 3 License No. DPR-47, page 3 License No. DPR-47, page 3 License No. DPR-55, page 3 License No. DPR-55, page 3 Technical Specifications Technical Specifications 5.0-7 5.0-7 5.0-8 5.0-8

Subsequent Renewed License No. DPR-38 Amendment 433

3.

This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 433, are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This subsequent renewed operating license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a)

Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b)

Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and Subsequent Renewed License No. DPR-47 Amendment No. 435

3.

This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 435 are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This subsequent renewed operating license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a)

Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b)

Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and Subsequent Renewed License No. DPR-55 Amendment No. 434

3.

This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 434 are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This subsequent renewed operating license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a)

Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b)

Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and

OCONEE UNITS 1, 2, & 3 5.0-7 Amendment Nos. 433, 435, & 434 Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program.

Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1.

sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2.

a determination that the change(s) do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b.

Shall become effective after the approval of the Plant Manager or Radiation Protection Manager; and

c.

Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Containment Leakage Rate Testing Program A program shall establish the leakage rate testing of the containment 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 Regulatory Guide 1.163, Revision 1 Performance-Based Containment Leak-Test Program, dated June 2023.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 59 psig. The containment design pressure is 59 psig.

Programs and Manuals 5.5 5.5 Programs and Manuals OCONEE UNITS 1, 2, & 3 5.0-8 Amendment Nos. 433, 435, & 434 The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containment air weight per day.

Leakage rate acceptance criterion is:

a.

Containment 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 Type B and C tests, and 0.75 La for Type A tests; The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies of 10 CFR 50, Appendix J.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR LICENSE AMENDMENT NOS. 433, 435, AND 434 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-38, DPR-47, AND DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287

1.0 INTRODUCTION

By application RA-25-0075 dated August 19, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25232A073), Duke Energy Carolinas, LLC (Duke Energy or the licensee) submitted a license amendment request (LAR) to the U.S.

Nuclear Regulatory Commission (NRC, The Commission) to request changes to the technical specifications (TSs) for the Oconee Nuclear Station, Units 1, 2, and 3 (Oconee, ONS).

The LAR proposes revising TS 5.5.2, Containment Leakage Rate Testing Program, to allow the following:

Increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years, Adopt an extension of the containment isolation valve (CIV) leakage rate testing frequency for Type C leakage rate testing of selected components, Adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2020, Containment System Leakage Testing Requirements, and Adopt 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 (ML12221A202).

2.0 REGULATORY EVALUATION

2.1

System Description

All three Units at ONS are Babcock and Wilcox pressurized water reactor plants.

In its letter dated August 19, 2025, the licensee stated, in part, that:

The containment consists of the reactor building (RB) structure, its steel liner, and the penetrations of this liner and structure. The containment is designed to contain radioactive material that may be released from the reactor core following a design basis loss of coolant accident (LOCA). Additionally, the containment provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The containment design includes ungrouted tendons where the cylinder wall is prestressed with a post tensioning system in the vertical and horizontal directions, and the dome roof is prestressed using a three way post tensioning system. The inside surface of the containment is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions.

2.2 Licensee Proposed Changes In its submittal, the licensee stated, in part, that:

The proposed change is a request to revise Technical Specifications (TS) 5.5.2 Containment Leakage Rate Testing Program, to allow the following:

Increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with the regulatory guidance specified in Regulatory Guide (RG) 1.163, Revision 1, Performance-Based Containment Leak-Test Program and Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A Adopt 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 75 months for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

Adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2020, Containment System Leakage Testing Requirements Adopt 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 to the TS contained herein would revise ONS TS 5.5.2, by replacing the references to the September 1995 revision of RG 1.163 with a reference to RG 1.163, Revision 1, dated June 2023 [ML23073A154]. This document will be used by ONS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Also, this license amendment request (LAR) proposes the following administrative changes to TS 5.5.2:

Deleting the information regarding the performance of containment visual inspections as required by RG 1.163, Regulatory Position C.3 as the containment inspections are presently addressed in TS SR 3.6.1.1.

Deleting the information regarding the performance of the next ONS Type A tests to be performed no later than November 29, 2026 for ONS 1, November 28, 2027 for ONS 2, and May 25, 2028 for ONS 3 as these Type A tests shall be scheduled in accordance with RG 1.163, Revision 1.

2.3 Applicable Regulatory Requirements The NRC staff considered the following regulations and guidance during its review of the proposed TS change.

The regulations in Title 10 of the Code of Federal Regulations (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. Appendix J to 10 CFR Part 50 specifies containment leakage testing requirements, including the types required to ensure the leaktight integrity of the primary reactor containment and systems and components, which penetrate the containment. In addition, Appendix J to 10 CFR Part 50 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 for meeting the requirements of the Appendix. The licensee adopted Option B of 10 CFR Part 50, Appendix J, for integrated (Type A) leakage rate testing with Amendment Nos. 310, 310, and 310 (Oconee, Units 1, 2, and 3, respectively) (ML003680348) and implemented Option B of 10 CFR Part 50, Appendix J with Amendment Nos. 375, 377, and 376 (Oconee, Units 1, 2, and 3, respectively) (ML11186A906) and is part of the Oconee current licensing basis.

The regulations in 10 CFR 50.55a, Codes and standards, contain the containment inservice inspection requirements, which, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leaktight and structural integrity of the containment during its service life.

The regulations in 10 CFR 50.36, Technical specifications, state that the TSs must include items in the following five specific categories: (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 regulations in 10 CFR 50.36(c)(5),

Administrative controls, state, in part, that:

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.

2.4 Regulatory Guidance NEI 94-01, Revision 3-A, provides methods for complying with the provisions of 10 CFR Part 50, Appendix J, Option B, 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 and guidance for extending Type C local leakage rate test (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 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 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.

Electric Power Research Institute (EPRI) Report No. 1009325, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A1, dated October 2008, provides a generic assessment of the risks associated with a permanent extension of the ILRT surveillance interval to 15 years, and a risk-informed methodology to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used, in combination with ILRT performance data and other considerations, to justify the extension of the ILRT surveillance interval. This is consistent with the guidance provided in RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256), and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, May 2011 (ML100910008), to support changes to surveillance test intervals.

RG 1.163, Revision 1, endorses NEI 94-01, Revision 3-A, as an acceptable method for implementing Option B of Appendix J to 10 CFR Part 50, subject to the regulatory positions listed in section C of the 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. RG 1.163, Revision 1, also endorses EPRI Report No.1009325, Revision 2-A, subject to the applicable regulatory positions listed in section C of the RG. In addition, RG 1.163, Revision 1, also endorses ANSI/ANS 56.8-2020, for acceptable industry standards on technical methods and techniques for performing Type A, B, and C tests.

RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities Revision 3 (ML20238B871) describes an acceptable approach for determining if a base PRA, in total or in the portions that are used to support an application, is sufficient to provide confidence and technical adequacy for regulatory decisionmaking.

3.0 TECHNICAL EVALUATION

3.1 Integrated Leak Rate Testing History (Type A Testing)

Oconee TS 5.5.2 specifies a maximum allowable containment leakage rate acceptance criterion (La) of 0.20 percent of the primary containment air weight per day at the calculated peak pressure, Pa. The peak containment internal pressure for a design basis loss-of-coolant-accident is 59 pounds per square inch gauge (psig). The containment leakage rate testing program acceptance criteria is less than or equal to 1.0 La.

There have been ten ILRTs per Unit performed on Oconee, Units 1, 2, and 3, and these tests have shown satisfactory leakage rate test results. The licensee provided the test results in 1 EPRI Report 1018243 is also identified as EPRI Report 1009325, Revision 2-A. This report is publicly available and can be found at www.epri.com by typing 1018243 in the search box.

Section 3.2.4, Integrated Leakage Rate Testing (ILRT) History, of the enclosure to the LAR.

The results are summarized in Tables 3.2.4-1 and 3.2.4-2 of the enclosure to the LAR.

The NRC staff reviewed the past ILRT results for Oconee and noted that there has been substantial margin maintained relative to the performance criterion. Since the last two Type A tests for Oconee had as found test results well within the current maximum allowable containment leakage rate specified in TS 5.5.2 of 0.20 weight-percent/day (1.0 La), there is sufficient basis for the Type A test frequency to be extended to 15 years in accordance with NEI 94-01, Revision 3-A and the regulatory positions in RG 1.163, Revision 1.

Based on the above, the NRC staff concludes that the Oconee ILRT test results provide reasonable assurance that containment overall leakage will be maintained below the design-basis leak rate, consistent with the requirements in TS 5.5.2, and will fulfill the requirements of 10 CFR Part 50, Appendix J, Option B, with the proposed test frequency of 15 years.

3.2 Type B and C Testing The Oconee 10 CFR Part 50, Appendix J, Type B and Type C leakage rate test program requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves within the scope of the program, as required by 10 CFR Part 50, Appendix J, Option B and TS 5.5.2.

The licensee states that in accordance with TS 5.5.2, the maximum allowable leakage rate for Type B and C testing is 0.6 Laor 60 percent, which equates to 212,402 standard cubic centimeters per minute (sccm), where La equals 354,444 sccm. In Section 3.4.6, Primary Containment Leakage Rate Testing Program - Type B and Type C Testing Program, of the enclosure to the LAR, the licensee provides a summary of LLRT as-found/as-left trends, which lists five previous LLRT tests for Unit 1, Unit 2, and Unit 3. These are listed in Tables 3.4.6-1, 3.4.6-2, and 3.4.6-3 in the enclosure to the LAR.

The results of the LLRTs have shown satisfactory leakage rate test with averages of approximately 6-9 percent of La, which satisfies the leakage rate test acceptance criteria of 60 percent of La.

Based on the NRC staffs review of the historical information provided in LAR Sections 3.4.6 and 3.4.7, Type B and Type C LLRT Program Implementation Review, the NRC staff noted that the licensee is adequately implementing the testing program in accordance with the requirements of 10 CFR Part 50, Appendix J, Option B performance-based testing program. In addition, the licensee has a corrective action program that appropriately addresses poor performing valves and penetrations.

The NRC staff finds that the licensee is effectively implementing the Oconee Type B and C leakage rate test program, as required by Option B of 10 CFR Part 50, Appendix J. Therefore, extending the containment isolation valve leakage rate testing (Type C) frequency from 60 months 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.

3.3 Containment Inspection 3.3.1 Nuclear Coatings Program and Protective Coating Monitoring and Maintenance Program In Section 3.4.1, Nuclear Coatings Program, of the enclosure to the LAR, the licensee described the ONS containment coatings program and its requirements for assessing the condition of Service Level I coatings inside the reactor containment building. These coatings are applied to surfaces where coating failure could adversely affect the operation of post-accident fluid systems, such as the emergency core cooling system (ECCS) and containment spray system, by introducing debris that could impair sump performance and safe shutdown.

The NRC staff reviewed the licensees description of the nuclear coatings program and the results of recent inspections provided in Section 3.4.1 of the enclosure to the LAR and Section 3.3.1 of the licensees submittal RA-23-0182 dated November 16, 2023 (ML23320A111). The NRC staff noted that the program includes condition assessments during each refueling outage, identification of degraded coatings, and scheduling of repairs or replacements. The NRC staff also noted that the licensee tracks the quantity of unqualified coatings to ensure that the total amount remains within the bounds assumed in the design basis analysis, with margin below administrative limits. The most recent inspections identified mostly minor damage, such as scratches, dings, and localized flaking of the topcoat, and no immediate corrective actions were required to maintain compliance with design and licensing basis requirements.

In Section 3.6.3, Protective Coating Monitoring and Maintenance, of the enclosure to the LAR, the licensee described its program for monitoring and maintaining Service Level I protective coatings inside containment. These coatings are applied to carbon steel and concrete surfaces where coating failure could adversely affect the operation of post-accident fluid systems, such as the ECCS, by introducing debris that could impair sump performance and safe shutdown.

The NRC staff reviewed the licensees description of the protective coating monitoring and maintenance program in Section 3.6.3 of the enclosure to the LAR. The staff noted that the program follows ASTM D5163-05a, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants, and includes coating selection, application, inspections, assessments, repairs, and maintenance. Qualified inspectors perform visual inspections of steel and concrete surfaces inside containment to ensure coating integrity. The program also controls the amount of degraded and unqualified coatings to keep them within design limits, supporting ECCS operability by preventing excessive debris. Therefore, the NRC staff finds that the program manages coating integrity sufficiently to minimize corrosion and maintain compliance with design and licensing requirements.

Based on the above, the NRC staff finds that the ONS nuclear coatings program and the protective coating monitoring and maintenance program are sufficient in monitoring the condition of Service Level I coatings inside containment, performing timely repairs, and ensuring that degraded or unqualified coatings remain within acceptable limits. Therefore, the NRC staff concludes that the licensee is sufficiently implementing both the nuclear coatings program and the protective coating monitoring and maintenance program.

3.3.2 Containment Inservice Inspection Program In Section 3.4.2, Containment Inservice Inspection (CISI) Plan of the enclosure to the LAR, the licensee described its CISI program for the fourth 10-year interval, which begins on July 15, 2024, and ends on July 14, 2034, has been developed in accordance with the requirements of 10 CFR 50.55a(g)(4)(ii). The licensee stated that the CISI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),

Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2019 Edition, as modified by the conditions in 10 CFR 50.55a.

3.3.2.1 ASME Section XI, Subsection IWE The NRC staff reviewed the information provided in Section 3.4.2 of the enclosure to the LAR and confirmed that the licensees CISI program for Class MC components is consistent with the requirements of ASME Section XI, Subsection IWE, as incorporated by reference in 10 CFR 50.55a. Each 10-year inspection interval consists of three examination periods, which are scheduled across three refueling outages, in accordance with ASME Code requirements. The staff verified that the containment inservice inspection intervals for the facility are properly established and meet the applicable regulatory and code requirements.

The NRC staff reviewed the information provided in Section 3.1.1 of the enclosure to the LAR regarding replacement of the steam generators requiring a construction opening in the reactor building shell wall. The LAR stated that portions of the containment liner plate and stiffeners removed during steam generator replacement were either reused or replaced with materials of the same grade as the existing structure. Fabrication was performed in accordance with ASME Section VIII, 1998 Edition with Addenda, and testing was conducted per ASME Section XI, Subsections IWE, IWL, and IWA, 1992 Edition with Addenda. Repairs complied with the original liner plate specifications.

In addition, the NRC staff reviewed the results of recent ASME Code,Section XI, Subsection IWE examinations presented in Section 3.4.4 of the enclosure to the LAR and confirmed that no indications of significant degradation were identified. The reported conditions were generally minor in nature. For example, some findings involved localized corrosion and pitting on the containment liner plate. In one case, ultrasonic testing (UT) measurements confirmed that the liner plate thickness at the deepest pit was approximately 0.225 inches, representing about a 10 percent reduction from the nominal 0.25-inch thickness. The licensee evaluated these conditions through engineering analysis and determined them to be acceptable.The NRC staff also noted that the licensee identified issues that did not meet acceptance criteria or indicated degradation, entered them into the sites corrective action program, and planned and implemented corrective actions.

Based on the above, the NRC staff concludes that the licensee has sufficiently implemented its ASME Section XI, Subsection IWE program for containment inservice inspections.

3.3.2.2. ASME Section XI, Subsection IWL The NRC staff reviewed the information provided in Section 3.4.2 of the enclosure to the LAR and confirmed that the licensees CISI program for ASME Class CC containment concrete components and the unbonded post-tensioning system is consistent with the requirements of ASME Section XI, Subsection IWL, as incorporated by reference in 10 CFR 50.55a. Each 10-year inspection interval consists of two examination periods, as required by the ASME Code.

The NRC staff verified that the containment inservice inspection intervals for the facility are properly established and meet the applicable regulatory and code requirements. The NRC staff noted that the licensee also plans to implement the requirements in 10 CFR 50.55a(b)(2)(viii)(H) and (b)(2)(viii)(I) for the examination of concrete containments.

The NRC staff reviewed the information provided in Section 3.1.1 of the enclosure to the LAR and finds that during the steam generator replacement, a temporary construction opening was created in the reactor building shell wall, requiring relaxation and later re-tensioning of affected tendons. New tendons installed were of the same type as existing, and replacement concrete was developed to ensure compatibility with the original structure. Augmented examinations of tendons impacted by these activities were performed in accordance with ASME Section XI, Subsection IWL-2521.2. Post-modification testing included pressurizing the containment to design pressure and performing a Type A Integrated Leak Rate Test to verify structural integrity.

In addition, the NRC staff reviewed the results of recent ASME Code,Section XI, Subsection IWL examinations presented in Sections 3.4.4 and 3.4.5 of the enclosure to the LAR. The NRC staff further reviewed the evaluation of non-conformances identified during recent ASME Section XI, Subsection IWL examinations and found that the reported conditions primarily involved low reserve alkalinity in some tendons, insufficient grease fill volumes, concrete cracks near tendon anchorages, and one instance of an undocumented broken wire. The licensee evaluated these conditions in accordance with IWL-3310, which requires assessment of items that do not meet acceptance standards in IWL-3100 or IWL-3200. The NRC staff noted that these conditions are minor in nature and do not impact the integrity or function of the containment pressure boundary. The NRC staff confirmed that augmented examinations of tendons impacted by the steam generator replacement did not identify any abnormal conditions affecting containment integrity. The NRC staff also noted that the licensee identified issues that did not meet acceptance criteria or indicated degradation, entered them into the sites corrective action program, and planned and implemented corrective actions.

Based on this review, the NRC staff concludes that the licensee has sufficiently implemented its ASME Section XI, Subsection IWL program for containment inservice inspections.

3.3.2.3 Supplemental Inspections Requirements In Section 3.4.3, Supplemental Inspection Requirements of the enclosure to the LAR, the licensee described its program for performing supplemental visual inspections of the Oconee containment structure. The licensee stated that general visual examinations will be performed in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsections IWE and IWL, as incorporated by reference in 10 CFR 50.55a. The inspections include the metallic liner, accessible concrete surfaces, penetrations, and post-tensioning system components of the containment structure.

The licensee also stated that general visual examinations conducted within 90 days prior to a Type A test may be used to satisfy the inspection requirements of 10 CFR 50, Appendix J.

Inspections are performed by qualified personnel, with all examination results and sign-offs overseen by the Responsible Individual, as defined in ASME Sections IWE-2330 and IWL-2330.

The Responsible Individual ensures development of examination procedures, personnel training, evaluation of inspection results, and review of any repair/replacement or pressure test procedures.

The NRC staff reviewed the licensees supplemental inspection program, including the inspection scope, frequency, timing relative to Type A tests, and personnel qualifications. The staff finds that the licensees program provides sufficient verification of both structural integrity and leak-tightness of the containment structure, consistent with 10 CFR 50, Appendix J, Options A and B, and Technical Specifications.

Based on this review, the NRC staff concludes that the licensees supplemental inspection program for the Oconee containment structure satisfies the applicable provisions of ASME Section XI, 10 CFR 50.55a, and 10 CFR 50, Appendix J.

3.3.3 Operating Experience In Section 3.5, Operating Experience, of the enclosure to the LAR, the licensee evaluated the following site-specific and industry events for the impact on the containment.

In Section 3.5.1, IN 1992-20, Inadequate Local Leak Rate Testing, of the enclosure to the LAR, the licensee evaluated the applicability of NRC Information Notice (IN) 92-20, Inadequate Local Leak Rate Testing, dated March 3, 1992 (ML031200473), to Oconee. The NRC issued IN 92-20 to inform licensees of issues associated with the local leak rate testing of two-ply stainless-steel bellows used on piping penetrations at certain facilities. The licensee concluded that IN 92-20 is not applicable to Oconee because only bellows assemblies are associated with fuel transfer tube penetrations that are welded to the containment liner, thereby eliminating a potential atmospheric release pathway.

In Section 3.5.2, IN 2010-12, Containment Liner Corrosion, of the enclosure to the LAR, the licensee evaluated the applicability of NRC Information Notice (IN) 2010-12, Containment Liner Corrosion, dated June 18, 2010 (ML100640449), to Oconee. The NRC issued IN 2010-12 to inform licensees of issues associated with corrosion of the steel reactor containment building liner. The licensee stated that IN 2010-12 is applicable to Oconee and its containment ASME Section XI, Subsections IWE and IWL programs and examination procedures contain steps to perform visual examinations for corrosion of the steel liner and identify liner bulge areas.

In Section 3.5.3, RIS 2016-07, Containment Shell or Liner Moisture Barrier Inspection, of the enclosure to the LAR, the licensee reviewed NRC Regulatory Issue Summary (RIS) 2016-07, Containment Shell or Liner Moisture Barrier Inspection, dated May 9, 2016 (ML16068A436),

for applicability to Oconees containment moisture barrier inspection program. The NRC issued RIS 2016-07 to reiterate the NRCs position to all licensees regarding the inservice inspection requirements for moisture barrier materials. In accordance with ASME Section XI, Table IWE-2500-1, Item E1.30, the NRC expects licensees to examine 100 percent of accessible moisture barriers during each inspection period. The licensee stated that RIS 2016-07 is applicable to Oconee and described the actions taken during outages 1EOC29, 2EOC28, and 3EOC29 to verify compliance. Specifically, the licensee: (1) identified all containment locations with metal-to-metal, metal-to-concrete, and concrete-to-concrete interfaces and documented the presence, type, and configuration of any associated moisture barriers or coatings; and (2) when moisture barriers were found to be missing or degraded, implemented appropriate corrective actions, such as installing new barriers, revising the Inservice Inspection (ISI) Plan, or providing technical justification. The licensee also ensured that examination procedures include clear guidance regarding the scope of inspections and applicable acceptance criteria for all moisture barriers. The NRC staff reviewed Table 3.4.2-2, Category E-A: Containment Surfaces, in Section 3.4.2 of the enclosure to the LAR and noted the licensee has committed to inspecting 100% of accessible moisture barriers during each inspection period within every inspection interval, consistent with the expectations of RIS 2016-07 and requirements of the ASME Code,Section XI.

In Section 3.5.4, ML24110A112, Duke Energys response to [Request for Additional Information] RAI questions, of the enclosure to the LAR, the licensee summarized its responses (ML24110A112) to two RAI in support of the LAR to Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program for a One-Time Extension of the Units 1, 2 and 3 Type A Leak Rate Test Frequency, dated November 16, 2023 (ML23320A111).

Based on the above, the NRC staff finds that the licensee has demonstrated that site-specific and industry operating experience, as well as lessons learned from previous NRC information requests, have been effectively incorporated into containment CISI programs.

3.4 Containment Accident Pressure on ECCS Performance In Section 3.1.3, Containment Overpressure on ECCS Performance, of the enclosure to the LAR, the licensee discussed the available net positive suction head (NPSH) for the Reactor Building Spray (RBS) in recirculation mode in a design-basis accident. The NRC staff uses the term containment accident pressure (CAP) for the pressure generated in the containment during an accident instead of containment overpressure. This section states:

ONS credits 0.44 psi of overpressure for calculation of available NPSH for the Reactor Building Spray (RBS) in recirculation mode from approximately 3000 to 30000 seconds in a design-basis accident. The Low-Pressure Injection (LPI) pumps do not require overpressure credit for NPSH. The RBS pumps are not credited in the ONS PRA, so there is no impact to the ILRT extension application due to the RBS pumps requirement for containment overpressure for adequate NPSH.

Since containment overpressure is not credited in the PRA models, no delta core damage frequency (CDF) estimate is required and there is no impact to the ILRT extension PRA evaluation.

Based on the above, the NRC staff finds that the licensee will maintain the current licensing basis NPSH analysis and no credit for CAP on ECCS performance is taken.

3.5 RG 1.163. Revision 1, Staff Regulatory Guidance RG 1.163, Revision 1 endorses NEI 94-01, Revision 3-A, which provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50, subject to the regulatory positions identified in section C of this RG. In LAR table 3.7.1-1, RG 1.163 Revision 1, Staff Regulatory Guidance, the licensee provided a response to each of these regulatory positions.

3.5.1 RG 1.163, Revision 1, Regulatory Position 1 Regulatory Position 1 of RG 1.163, Revision 1, states in part that:

ANSI/ANS-56.8-2020 as approved by this RG may be used, and for calculating the Type A leakage rate, the licensee should use the performance leakage rate definition in NEI 94-01, Revision 3-A.

The licensees response to Regulatory Position 1 states:

ONS will utilize ANSI/ANS 56.8-2020.

ONS will utilize the definition in NEI 94-01 Revision 3-A, Section 5.0.

The licensees response is consistent with RG 1.163, Revision 1. Therefore, the NRC staff concludes that the licensee sufficiently addressed Regulatory Position 1.

3.5.2 RG 1.163, Revision 1, Regulatory Position 2 Regulatory Position 2 of RG 1.163, Revision 1, states:

The licensee should submit a schedule of containment inspections to be performed before and between Type A tests as part of the LAR submittal for Type A test interval extension.

The licensees response to Regulatory Position 2 states:

Reference Section 3.4.2 of this submittal.

In Section 3.4.2 of the enclosure to the LAR, the licensee provided schedules for the ASME Code,Section XI, Subsection IWE examinations for Examination Categories E-A, E-C and E-G components at Oconee. For the fourth 10-year CISI interval, the IWE examination schedules for Units 1, 2, and 3 are summarized in Tables 3.4.2-27, 3.4.2-29, and 3.4.2-31 of the enclosure to the LAR, respectively. For the fifth 10-year CISI interval, the IWE examination schedules for Units 1, 2, and 3 are summarized in Tables 3.4.2-33, 3.4.2-35, and 3.4.2-37 of the enclosure to the LAR, respectively. The licensee also provided schedules for the ASME Code,Section XI, Subsection IWL examinations at Oconee for Examination Category L-A and L-B components.

For the fourth 10-year CISI interval, the IWL examination schedules for Units 1, 2, and 3 are summarized in Tables 3.4.2-28, 3.4.2-30, and 3.4.2-32 of the enclosure to the LAR, respectively. For the fifth CISI interval, the IWL examination schedules for Units 1, 2, and 3 are summarized in Table 3.4.2-34, 3.4.2-36, and 3.4.2-38 of the enclosure to the LAR, respectively.

Containment concrete is examined every 5 years as specified in ASME Code,Section XI, paragraph IWL-2410.

The licensees response is consistent with RG 1.163, Revision 1. Therefore, the NRC staff concludes that the licensee sufficiently addressed Regulatory Position 2.

3.5.3 RG 1.163, Revision 1, Regulatory Position 3 Regulatory Position 3 of RG 1.163, Revision 1, states in part that:

The LAR should address the areas of the containment structure potentially subject to degradation.

The licensees response to Regulatory Position 1 states:

Reference Section 3.4.2 of this submittal.

As described in Section 3.4.2 of the enclosure to the LAR, Oconees CISI program was developed in accordance with the requirements of the ASME Code,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a. In Section 3.6, License Renewal Aging Management, of the enclosure to the LAR, the licensee described the license renewal Aging Management Programs (AMPs), consisting of the ASME Section XI, Subsection IWE AMP and the ASME Section XI, Subsection IWL AMP. These programs are credited with managing age-related degradation of the containment pressure-retaining boundary, including the concrete structures and steel liner plate.

The NRC staff reviewed the CISI program and associated AMPs in Sections 3.4.2 and 3.6 of the enclosure to the LAR. The NRC staff finds that the licensee described the implementation of ASME Section XI, Subsections IWE and IWL, in sufficient detail to address areas of the containment structure potentially subject to degradation. These include containment pressure-retaining boundary components, as well as accessible and inaccessible areas of the containment structures. The NRC staff also finds that the program includes inspections of degradation-susceptible inaccessible areas using appropriate examination techniques consistent with ASME Code requirements and applicable guidance. Based on its review, the NRC staff concludes that the licensee has provided an acceptable level of information demonstrating that the CISI program and associated AMPs will manage age-related degradation of the containment structures during the period of extended operation.

The licensees response is consistent with RG 1.163, Revision 1. Therefore, the NRC staff concludes that the licensee sufficiently addressed Regulatory Position 3.

3.5.4 RG 1.163, Revision 1, Regulatory Position 4 Regulatory Position 4 of RG 1.163, Revision 1, states, in part:

As part of the LAR submittal, the licensee should provide information about any tests and inspections following major modifications to the containment structure, as applicable.

The licensees response to Regulatory Position 1 states:

Steam Generator replacements were conducted using construction openings in each containment followed by the performance of ILRTs following restoration. Reference Section 3.1.1 of this submittal.

In Section 3.1.1 of the enclosure to the LAR, the licensee stated that the replacement of the steam generators at Oconee required creating a construction opening in the shell wall of the reactor buildings. Following completion of the steam generator replacement activities, a Type A ILRT was performed in accordance with the requirements of 10 CFR Part 50, Appendix J. The results of the Type A tests, provided in Tables 3.2.4-1 and 3.2.4-2 of the enclosure to the LAR, indicate that the ILRTs conducted in December 2003 (Unit 1), May 2004 (Unit 2), and December 2004 (Unit 3) yielded containment air weight leakage rates of 0.097334 percent, 0.093723 percent, and 0.07075 percent per day, respectively. These measured leakage rates are below the ILRT acceptance criterion of 0.1875 percent per day.

The NRC staff noted that the licensee performed the required leakage rate tests following each major containment modification associated with the steam generator replacement activities and that the results of each test met the applicable acceptance criteria. Based on the above, the NRC staff finds that the licensees actions are consistent with the NRC staffs position regarding post-repair pressure testing following major containment modifications.

The licensees response is consistent with RG 1.163, Revision 1. Therefore, the NRC staff concludes that the licensee sufficiently addressed Regulatory Position 4.

3.5.5 RG 1.163, Revision 1, Regulatory Position 5 Regulatory Position 5 of RG 1.163, Revision 1, states:

The normal Type A test interval should be less than 15 years. If a licensee desires to use the provision of Section 9.1 of NEI 94-01, Revision 3-A, related to extending the ILRT interval beyond 15 years, the licensee should demonstrate in a LAR that the extension is necessary due to an unforeseen emergent condition (see Regulatory Issue Summary 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50, dated December 8, 2008 [Reference 23].

The licensees response to Regulatory Position 5 states:

ONS will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1.

In accordance with the requirements of NEI 94-01, Revision 3-A, Section 9.1, Duke Energy will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

The licensees response is consistent with RG 1.163, Revision 1. Therefore, the NRC staff concludes that the licensee sufficiently addressed Regulatory Position 5.

3.5.6 RG 1.163, Revision 1, Regulatory Position 6 Regulatory Position 6, of RG 1.163, Revision 1, addresses new reactor plants licensed under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

Regulatory Position 6 is not applicable to Oconee because it was not licensed under 10 CFR Part 52. Oconee was licensed under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.

3.5.7 Conclusion Related to the Regulatory Positions Listed in RG 1.163, Revision 1 The NRC staff evaluated the first six regulatory positions listed in RG 1.163, Revision 1, section C, and determined that the licensee adequately addressed each of them. The remaining regulatory positions listed in RG 1.163, Revision 1, Section C are addressed in section 3.7 of this SE. Therefore, the NRC staff finds it acceptable for the licensee to adopt RG 1.163, Revision 1, as the implementation document listed in TS 5.5.2.

3.6 NEI 94-01, Revision 3-A, Conditions The NRC published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012. In section 3.7.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A, of the enclosure to the LAR, the licensee provided a response to each of these conditions.

3.6.1 NEI 94-01, Revision 3-A, Condition 1 Condition 1 identifies three issues that are required to be addressed:

(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; (2) A corrective action plan is to be developed to restore the margin to an acceptable level; and (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 states:

The post-outage report shall include the margin between the Type B and Type C Minimum Pathway Leak 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 states:

When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the ONS, Units 1, 2, and 3 administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the ONS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

The licensees response to Condition 1, Issue 3 states:

ONS, Units 1, 2, and 3 will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

The NRC staff reviewed the licensees responses and finds that each of the three issues has been sufficiently addressed, and therefore Condition 1 of the NEI 94-01, Revision 3-A, SE has been sufficiently addressed.

3.6.2 NEI 94-01, Revision 3-A, Condition 2 Condition 2 identifies two issues that are required to be addressed:

(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.

(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 & 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 the penetrations.

The licensee provided a response to Condition 2 in LAR Section 3.7.2. For Condition 2, Issue 1, the licensee will conservatively apply a potential leakage understatement adjustment factor. For Condition 2, Issue 2, the licensee will monitor and trend the test program and prepare a post-outage report presenting test results.

The NRC staff has reviewed the licensees responses for Issues (1) and (2) to Condition 2 of NEI 94-01, Revision 3-A. The licensees responses indicate that the licensees actions will be consistent with the guidance of NEI 94-01, Revision 3-A. The NRC staff notes that revised guidance contained in NEI 94-01, Revision 3-A, Section 11.3.2 Programmatic Controls, and section 12.1 Report Requirements, reflects the NRC staffs SE input pertaining to both Issues (1) and (2). The NRC staff concludes that the licensee has accepted all the issues of Condition 2, and that the licensee has established programs for Oconee to comply with these requirements; therefore, the licensee has adequately addressed Condition 2.

3.6.3 NEI 94-01, Revision 3-A, Conditions Conclusion Based on the above evaluation of each condition, the NRC staff determined that the licensee has sufficiently addressed the conditions in Section 4.0 of the NRC SE of NEI 94-01, Revision 3.

Therefore, the NRC staff finds it acceptable for the licensee to adopt RG 1.163, Revision 1, which endorses NEI 94-01, Revision 3-A, as the implementation document listed in Oconee TS 5.5.2.

3.7 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals In its submittal dated August 19, 2025, the licensee states:

A plant specific risk assessment was conducted to support this proposed change. This risk assessment followed the guidelines of Nuclear Regulatory Commission (NRC) RG 1.174, Revision 3 and RG 1.200, Revision 3... The risk assessment concluded that increasing the ILRT frequency on a permanent basis from a one-in-ten-year frequency to a one-in-fifteen-year frequency is considered to be small since it represents a small change to the ONS risk profile.

In the enclosure to the LAR, the licensee provided a plant specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years.

In the enclosure to the LAR, the licensee described the risk insights that form the basis for its proposed change and support the deterministic evaluation of the proposed permanent change to the integrated leakage rate test (ILRT) time extension. The licensee stated that a plant specific risk assessment was conducted to support this proposed change and that the risk assessment followed the guidelines of RG 1.174, Revision 3 (ML17317A256) and RG 1.200, Revision 3. In the enclosure to the LAR, the licensee addressed Regulatory Positions 7 through 10 of RG 1.163, Revision 1 (ML23073A154), which are consistent with the four conditions imposed by the NRC staff in Section 4.2 of the NRC SE dated June 25, 2008, for the use of EPRI Report 1009325, Revision 2 (ML072970208). A summary of how each of the four conditions are met is provided in sections 3.7.1 through 3.7.4 below.

3.7.1 PRA Quality - RG 1.163, Revision 1, Regulatory Position 7 The first condition stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the guidance in RG 1.200 relevant to the ILRT extension application. RG 1.200 describes one acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results such that the PRA can be used in regulatory decision-making for light-water reactors.

In Section 3.3.2 of the LAR enclosure, PRA Acceptability, the licensee states that the internal events PRA model was subject to a full scope peer review in October 2013. A formal facts and observations closure review was performed per NEI 05-04 (ML041480432) and ANS/ASME PRA Standard, using the NRC-accepted process documented in the NEI letter to the NRC Final Revision of Appendix X to NEI 05-04/07-12/12-13 (ML17079A427). The licensee stated that all Level 1 model finding Facts and Observations (F&Os) were closed and all surveillance requirements (SRs) were met. The NRC staff concluded that the internal events (including internal flooding) PRA model was appropriately peer-reviewed, consistent with RG 1.200, Revision 3, and thus adequate to assess the changes to ILRT frequencies.

The licensees fire PRA was subject to a full scope peer review using the peer review process in 2012. After updating the fire PRA, a focused-scope peer review was performed in 2019 which resulted in most of the F&Os being closed. After additional resolution work was incorporated into the Fire PRA, another closure review was performed in 2020 which closed the remaining F&Os. As a result, all associated supporting requirements were met at Capability Category II or better and there are no remaining open-finding-level F&Os. Therefore, no F&Os associated with the fire PRA were provided in the LAR. The NRC staff concluded that the fire PRA model was appropriately peer-reviewed, consistent with RG 1.200, Revision 3, and thus sufficient to assess the changes to ILRT frequencies.

The licensees seismic PRA model was subject to a full scope review in June 2018 against the ASME/ANS RA-Sb-2013. The licensee mentioned an F&O closure review was performed on select F&Os. Several F&Os were confirmed closed, while others were partially closed with documentation updates required, and other F&Os that were not reviewed by the closure team.

The 25 F&Os that remain open are described and dispositioned in Attachment 2 of the LAR.

The NRC staff reviewed the dispositions of the seismic PRA F&Os and determined that the licensee properly dispositioned the F&Os for this application. Of the remaining F&Os that are open, the disposition provided justification for why they were not resolved and was acceptable for this application. For most of the F&Os the licensee discussed that the overall risk will be small and/or have little to no impact on this application.

Based on the above, the NRC staff concludes that the PRA models used by the licensee are of sufficient quality to support the evaluation of changes to ILRT frequencies. Accordingly, the NRC staff concludes that Regulatory Position 7 has been satisfactorily addressed.

3.7.2 Estimated Risk Increase - RG 1.163, Revision 1, Regulatory Position 8 The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, consistent with the guidance in RG 1.174 and the clarification provided in the NRC SE for EPRI Report No. 1009325. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem (roentgen equivalent man) per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points.

The licensee reported the results of the plant-specific risk assessment in section 3.3.3, Summary of Plant-Specific Risk Assessment Results, of the enclosure to the LAR. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency under 10 CFR Part 50 Appendix J, Option A) to one test in 15 years. The licensee drew the following conclusions from its analysis associated with extending the Type A ILRT frequency:

1. Large early release frequency (LERF) is the relevant risk metric for this LAR.

RG 1.174 defines very small changes in risk as resulting in an increase of LERF less than 1.0E-7 per reactor year and considers a small change in LERF to be between 1E-7 and 1E-6 per reactor year with a total LERF less than 1E-5 per reactor year. The increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 9.96E-07/year, 9.73E-07/year, and 8.93E-07/year respectively, for Units 1, 2, and 3. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 4.15E-07/year, 4.06E-07/year, and 3.72E-07/year respectively, for Units 1, 2, and 3. Therefore, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174.

2. The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0076, 0.0071, and 0.0072 person-rem/year respectively, for Units 1, 2, and 3. This is below the acceptance criteria provided in the NRC SE for EPRI Report No. 1009325, Revision 2, which states that a small total population dose is defined as an increase of 1.0 person-rem/year, or 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. Thus, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

3. The increase in the CCFP due to the change in the test frequency from 3 in 10 years to 1 in 15 years is 0.89 percent. This estimated increase is below the acceptance guideline in accordance with NEI 94-01 which states that the increase in CCFP of 1.5 percent is small, and thus supportive of the proposed change.

Based on the review of the licensees risk assessment results, the NRC staff concludes that the increase in LERF is small and consistent with the acceptance guidelines of RG 1.174, and the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are small. The defense-in-depth philosophy is maintained as the independence of barriers will not be degraded because of the requested change, and the use of the quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.

Accordingly, the NRC staff concludes that Regulatory Position 8 has been sufficiently addressed.

3.7.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case - RG 1.163, Revision 1, Regulatory Position 9 The third condition stipulates that for the methodology in EPRI Report No. 1009325 to be acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by the licensees shall be 100 La instead of 35 La. As noted by the licensee in table 3.3.1-1 of the enclosure to the LAR, 100 La was used as the representative containment leakage for case 3b sequences in the plant-specific risk assessment, based on the guidance provided in EPRI report No. 1009325, Revision 2-A. Accordingly, the NRC staff concludes that Regulatory Position 9 has been sufficiently addressed.

3.7.4 Containment Overpressure is Relied Upon for ECCS Performance - RG 1.163, Revision 1, Regulatory Position 10 The fourth condition stipulates that, in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. In table 3.3.1-1 of the enclosure to the LAR, the licensee stated that Oconee credits 0.44 psi of overpressure of available net positive suction head (NPSH) for ECCS design basis accidents. Low Pressure Injection pumps are not required in support of ECCS performance to mitigate design-basis accidents. The licensee stated that the RBS pumps are not credited in its PRA, so there is no impact to this application due to the RBS pumps requirement for containment overpressure for adequate NPSH. The licensee further stated that since containment overpressure is not credited in its PRA models, no change in CDF (CDF) estimate is required and there is no impact to this application. Accordingly, the NRC staff concludes that Regulatory Position 10 has been satisfactorily addressed.

3.8 NRC Staff Conclusion

Based on the above, the NRC staff finds that the licensee has sufficiently implemented its existing primary containment leakage rate testing program consisting of ILRT and LLRT. Results of the recent ILRTs and of the LLRTs combined totals demonstrate acceptable performance and support a regulatory conclusion that the structural and leak-tight integrity of the primary containment is sufficiently managed and will continue to be periodically monitored and maintained effectively with the proposed changes. The NRC staff finds that the licensee has addressed the NRC regulatory positions to demonstrate acceptability of adopting RG 1.163 Revision 1, and the limitations and conditions identified in NEI 94-01, Revision 3A. The NRC staff also finds that the PRA used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequency. Therefore, the NRC staff concludes that the proposed changes to Oconee TS 5.5.2 regarding the containment leakage rate testing program are acceptable and continue to meet 10 CFR 50.36(c)(5).

Additionally, the licensee proposed deletion of the information regarding the performance of the next Type A tests under Oconee TS 5.5.2 scheduled November 29, 2026, for Unit 1 and November 28, 2027, for Unit 2, and May 25, 2028, for Unit 3. These activities will be scheduled in accordance with RG 1.163 Revision 1. The licensee also intends to delete information regarding the performance of containment visual inspections as required as it is already addressed in TS SR 3.6.1.1. Therefore, the NRC staff finds the proposed deletions under Oconee TS 5.5.2 acceptable because the deleted information is not required under 10 CFR 50.36(c)(5).

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendments on February 13, 2026. On February 13, 2026, the State official confirmed that the State of South Carolina had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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 published in the Federal Register on November 25, 2025 (90 FR 53396), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the 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:

B. Lee, NRR D. Ju, NRR C. Ashley, NRR N. Hansing, NRR G. Wang, NRR R. Atienza, NRR Date of Issuance: February 27, 2026

ML26027A354

• via eConcurrence NRR-058 OFFICE NRR/DORL/LPL2-1/PM*

NRR/DORL/LPL2-1/LA*

NRR/DSS/STSB/BC*

NAME SWilliams KZeleznock SMehta DATE 2/18/2026 2/20/2026 2/25/2026 OFFICE NRR/DSS/SNSB/BC*

NRR/DSS/SCPB/BC*

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NAME NDifrancesco MValentin SBailey DATE 2/24/2026 2/24/2026 2/21/2026 OFFICE NRR/DEX/ESEB/BC*

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NAME ITseng EDavidson MMarkley DATE 2/23/2026 2/25/2026 2/27/2026 OFFICE NRR/DORL/LPL2-1/PM*

NAME SWilliams DATE 2/27/2026