ML25063A335

Issuance of Amendment Nos. 248 and 250 Extension of Type a and Type C Leak Rate Test Frequencies

ML25063A335
Person / Time
Site:	Diablo Canyon
Issue date:	03/27/2025
From:	Samson Lee Plant Licensing Branch IV
To:	Gerfen P Pacific Gas & Electric Co
Lee S, 301-415-3158
References
EPID L-2024-LLA-0106
Download: ML25063A335 (30)

Text

March 27, 2025 Paula Gerfen Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424

SUBJECT:

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 248 AND 250 RE: EXTENSION OF TYPE A AND TYPE C LEAK RATE TEST FREQUENCIES (EPID L-2024-LLA-0106)

Dear Paula Gerfen:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 248 to Facility Operating License No. DPR-80 and Amendment No. 250 to Facility Operating License No. DPR-82 for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated July 31, 2024, as supplemented by letter dated February 12, 2025.

The amendments revise TS 5.5.16, Containment Leakage Rate Testing Program, for permanent extension of Type A and Type C leak rate test frequencies.

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

Sincerely,

/RA/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 248 to DPR-80

2. Amendment No. 250 to DPR-82

3. Safety Evaluation cc: Listserv

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 248 License No. DPR-80

1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated July 31, 2024, as supplemented by letter dated February 12, 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 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 248 are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-80 and the Technical Specifications Date of Issuance: March 27, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.03.27 16:33:56 -04'00'

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 250 License No. DPR-82

1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated July 31, 2024, as supplemented by letter dated February 12, 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 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:

(2)

Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 250, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-82 and the Technical Specifications Date of Issuance: March 27, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.03.27 16:35:23 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 248 TO FACILITY OPERATING LICENSE NO. DPR-80 AND LICENSE AMENDMENT NO. 250 TO FACILITY OPERATING LICENSE NO. DPR-82 DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of Facility Operating License Nos. DPR-80 and DPR-82, 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.

Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 5.0-16 5.0-16

Amendment No. 248 (4)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This License shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 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 The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 248 are hereby incorporated in the license.

Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Companys Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:

a.

Elimination of any test identified in Section 14 of PG&Es Final Safety Analysis Report as amended as being essential;

Amendment No. 250 (4)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This License shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 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 The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 250, are hereby incorporated in the license.

Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Initial Test Program (SSER 31, Section 4.4.1)

Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 248 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 250 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By application dated July 31, 2024 (Reference 1), as supplemented by letter dated February 12, 2025 (Reference 2), Pacific Gas and Electric Company (PG&E, the licensee) requested changes to the technical specifications (TSs) for the Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon or DCPP).

The license amendment request (LAR) proposes changes to Diablo Canyon TS 5.5.16, Containment Leakage Rate Testing Program, to allow for the permanent extension of the Type A Integrated Leak Rate Testing (ILRT) and Type C Leak Rate Testing frequencies based on the guidance in Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July, 2012 (Reference 3), as endorsed by Regulatory Guide (RG) 1.163, Revision 1, Performance-Based Containment Leak-Test Program, dated June, 2023 (Reference 4).

The supplemental letter dated February 12, 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 or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 1, 2024 (89 FR 79970).

2.0 REGULATORY EVALUATION

2.1 Description of Containment In section 3.1.1, Description of Primary Containment System, of the enclosure to the LAR, the licensee stated, in part:

The reactor containment building for each unit is a steel-lined, reinforced concrete cylindrical building with a dome roof that completely encloses the reactor and reactor coolant system (RCS). It ensures that essentially no leakage

of radioactive materials to the environment would result even if gross failure of the RCS were to occur simultaneously with a design basis earthquake.

2.2 Licensees Proposed Changes The licensees proposed change would revise Diablo Canyon TS 5.5.16, Containment Leakage Rate Testing Program, by replacing the reference to RG 1.163, Revision 0, Performance-Based Containment Leak-Test Program (Reference 5) with a reference to RG 1.163, Revision 1, as the document used to implement the performance-based containment leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

The proposed change would also delete the third exception under Diablo Canyon TS 5.5.16 for the 15-year Type A test interval beginning May 4, 1994, for Unit 1 and April 30, 1993, for Unit 2.

The exception was for activities that have already taken place and are no longer applicable.

In addition, the licensee will adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2020, Containment System Leakage Testing Requirements for technical methods and techniques when performing Type A, B, and C tests (Reference 6).

2.3 Regulatory Requirements 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) and local (Types B and C) leakage rate testing with Amendment Nos. 110 and 109 (Diablo Canyon, Units 1 and 2, respectively) (Reference 7) and is part of the Diablo Canyon 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:

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 (Reference 8).

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, (Reference 9) 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 (Reference 10), and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications, May 2011 (Reference 11), to support changes to surveillance test intervals.

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 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.

3.0 TECHNICAL EVALUATION

3.1 Integrated Leak Rate Testing History (Type A Testing)

Diablo Canyon TS 5.5.16 specifies a maximum allowable containment leakage rate acceptance criterion (La) of 0.10 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 43.5 pounds per square inch gauge (psig). The containment leakage rate testing program acceptance criteria is less than or equal to 1.0 La.

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.

There have been eight ILRTs performed on Diablo Canyon, Unit 1, and six ILRTs performed on Diablo Canyon, Unit 2, to date, and these tests have shown satisfactory leakage rate test results. The licensee provided the test results in 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 Diablo Canyon and noted that there has been substantial margin maintained relative to the performance criterion. Since the last two Type A tests for Diablo Canyon had as found test results well within the current maximum allowable containment leakage rate specified in TS 5.5.16 of 0.10 weight-percent/day (1.0 La), the Type A test frequency can 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 Diablo Canyon 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 5.5.16, and will fulfill the requirements of 10 CFR Part 50, Appendix J, Option B, with the test frequency of 15 years.

3.2 Type B and C Testing The Diablo Canyon 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 5.5.16.

The licensee states that in accordance with TS 5.5.16, the maximum allowable leakage rate for Type B and C testing is 0.6 Laor 60 percent, which equates to 119,143 standard cubic centimeters per minute (sccm), where La equals 198,581 sccm. In section 3.4.4, 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 six previous LLRT tests for Unit 1 and Unit 2. These are listed in tables 3.4.4-1 and 3.4.4-2 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 3.4.4 and 3.4.5, 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 Diablo Canyon 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 Coating Quality Monitoring Program In LAR section 3.4.1, Coating Quality Monitoring Program, the licensee described the protective coating program applied to all containment coatings that are performed during every refueling outage (RFO). The licensees program provides a common approach in controlling, applying, maintaining, and periodically assessing containment coatings where the coating failure could adversely affect the operation of the emergency core cooling system (ECCS) by clogging the suction strainers, which could possibly impair safe shutdown, and ensures deficient coatings are identified, evaluated, and repaired within a reasonable time. The licensee also discussed the recent coating inspection results and described the tracking process to quantify unqualified coatings to ensure that the documented quantity does not exceed the postulated maximum allowable quantity in the containment building. In the enclosure to the LAR, section 3.6.2, Protective Coating Monitoring and Maintenance Program AMP [Aging Management Program],

the licensee described the license renewal AMP that is used to manage the degradation of the Service Level I coatings subjected to indoor air in the containment building.

3.3.2 Containment Inservice Inspection Program In section 3.4.2, Containment Inservice Inspection Plan of the enclosure to the LAR, the licensee described its containment inservice inspection (CISI) program developed in accordance with the requirements in 10 CFR 50.55a. Visual inspections of the concrete are conducted in accordance 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, subsection IWL, while inspections of the liner are conducted in accordance with ASME Code,Section XI, subsection IWE. The licensee stated that its third interval, starting May 9, 2018, CISI program complies with ASME Code Section XI, subsections IWE and IWL, 2007 Edition with 2008 Addenda.

In section 3.4.2, of the enclosure to the LAR, the licensee presented the results of recent visual inspections for ASME Code,Section XI, subsection IWE examination during Unit 1 RFOs (1R20, 1R22, and 1R24), and Unit 2 RFOs (2R20, 2R21, and 2R22). The licensee evaluated the ASME Code,Section XI, subsection IWE inspection results and showed one reportable indication during 1R20, which is the abnormality of containment hatch bolt on Location No. 41, and the affected bolting was replaced during the same RFO. The licensee also presented the results of recent visual inspections for ASME Code,Section XI, subsection IWL examination in 2021 for both units, and concluded that containment concrete was structurally sound, no repairs were needed, and the containments continue to perform their intended function.

In section 3.4.3, Supplemental Inspection Requirements of the enclosure to the LAR, the licensee stated that supplemental inspections will not be required. Rather, inspections of the exterior containment concrete surfaces and the steel liner plate inside containment will be conducted in accordance with Diablo Canyon TS 5.5.16 as modified by Amendments Nos. 197 and 198 (Reference 12) by adding the following exceptions, which modify compliance with RG 1.163:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by

the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI code, Subsection IWE, except where relief has been authorized by the NRC.

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:

NRC Information Notice (IN) 1989-79, Degraded Coatings and Corrosion of Steel Containment Vessels (Reference 13)

NRC IN 1992-20, Inadequate Local Leak Rate Testing (Reference 14)

NRC IN 1997-10, Liner Plate Corrosion in Concrete Containments (Reference 15)

NRC IN 2004-09, Corrosion of Steel Containment and Containment Liner (Reference

16)

NRC IN 2010-12, Containment Liner Corrosion (Reference 17)

NRC IN 2014-07, Degradation of Leak Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner (Reference 18)

Regulatory Issue Summary (RIS) 2016-07, Containment Shell or Liner Moisture Barrier Inspection (Reference 19)

Diablo Canyon Power Plant, Unit 1 - NRC inspection report 05000275/2023011 (Reference 20).

The licensee provided the results of their review of the applicable INs, RIS, and NRC inspection report and demonstrated how such information is used to inform their programs for maintaining the overall containment integrity.

The NRC staff reviewed sections 3.5.1 through 3.5.8 of the enclosure to the LAR and finds that the licensee acceptably addressed the relevant regulatory requirements, guidance, and operating experience described above through inspection and aging management programs.

Therefore, the licensees CISI program provides reasonable assurance that the containment will maintain its capability to perform its safety-related function.

3.4 Containment Accident Pressure on ECCS Performance In section 3.1.2, Containment Overpressure on ECCS Performance, of the enclosure to the LAR, the licensee discussed the available net positive suction head (NPSH) for the ECCS pumps. The NRC staff uses the term containment accident pressure (CAP) for the pressure

generated in the containment during an accident instead of containment overpressure. The NPSH for the residual heat removal (RHR) pumps and the NPSH for the safety injection (SI) pumps and centrifugal charging pumps (CCP1 and CCP2), were evaluated for both the injection and recirculation modes of operation for the design-basis accident.

For the RHR pumps, the recirculation operation gives the limiting NPSH requirement based on all pumps (i.e. both RHR pumps, CCP1, CCP2, both SI pumps, and both containment spray system pumps) operating at the maximum design (run out) flow rates. In addition to considering the static head and suction line pressure drop, the calculation of available NPSH in the recirculation mode for the RHR pumps assumes that the vapor pressure of the liquid in the sump equals containment pressure. This assumption ensures that the actual available NPSH is always greater than the calculated NPSH.

For the SI pumps and CCP1 and CCP2, the end of the injection mode of operation gives the limiting NPSH available. The limiting NPSH was determined from the elevation head and vapor pressure of the water in the refueling water storage tank, which is at atmospheric pressure, and the pressure drop in the suction piping from the tank to the pumps. The NPSH evaluation is based on all pumps operating at the maximum design flow rates. Following switchover to the recirculation mode, adequate NPSH is supplied from the containment recirculation sump by the booster action of the RHR pumps.

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 (Reference 4), endorses NEI 94-01, Revision 3-A (Reference 3), 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:

DCPP will utilize ANSI/ANS 56.8-2020.

DCPP 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 adequately 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 licensee provided a schedule for the subsection IWE examinations of ASME Code,Section XI at Diablo Canyon for Examination Categories E-A, E-C and E-G components. The IWE examination schedule is summarized in tables 3.4.2-10 and 3.4.2-11 for Unit 1, DCPP Unit 1 IWE Examination Schedule, and tables 3.4.2-12 and 3.4.2-13 for Unit 2, DCPP Unit 2 IWE Examination Schedule, of the enclosure to the LAR. The schedule is detailed and organized by planned outage and period within the 10-year interval.

The licensee also provided a schedule for the subsection IWL examinations of ASME Code,Section XI at Diablo Canyon for Examination Category L-A, Containment Concrete Shell.

The IWL examination schedule is summarized in table 3.4.2-14, DCPP Unit 1 and 2 IWL Examination Schedule, of the enclosure to the LAR. Containment concrete is examined every 5 years as specified in ASME Code,Section XI, paragraph IWL-2410.

The NRC staff reviewed tables 3.4.2-10 to 3.4.2-14 of the enclosure to the LAR and finds that RG 1.163 Revision 1, Regulatory Position 2 has been met.

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

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

As described in section 3.4.2 of the enclosure to the LAR, Diablo Canyons CISI program is developed in accordance with the requirements of ASME Code Section XI, subsections IWE and IWL, as required by 10 CFR 50.55a. In section 3.6.1, Primary Containment Inservice Inspection Program, of the LAR enclosure, the licensee described the license renewal AMP, consisting of ASME Section XI, subsection IWE AMP and ASME Section XI, subsection IWL AMP, that is used to manage the degradation of the containment concrete and liner plate.

The NRC staff reviewed the CISI program, ASME Section XI, subsection IWE AMP, and ASME Section XI, subsection IWL AMP (References 21 and 22). The NRC staff finds that the licensee provided an acceptable level of information regarding the implementation of ASME Section XI, subsections IWE and IWL, that exhibit and/or would indicate the presence of potential degraded conditions in the accessible and inaccessible areas of the containment concrete and liner plate.

Therefore, the NRC staff finds that RG 1.163 Revision 1, Regulatory Position 3 has been met.

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.

In table 3.7.1-1, No. 4, of the enclosure to the LAR, the licensee stated that There are no major modifications planned.

Based on the information above, the NRC staff finds that RG 1.163 Revision 1, Regulatory Position 4 is not applicable.

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:

DCPP 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, DCPP 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 adequately 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 Diablo Canyon because it was not licensed under 10 CFR Part 52. Diablo Canyon 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 In this section of the SE, 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 5.5.16.

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 (Reference 8). 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 DCPP, Unit 1 and 2 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 DCPP 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:

DCPP, Units 1 and 2 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 satisfactorily addressed, and therefore Condition 1 of the NEI 94-01 Revision 3-A SE has been satisfactorily 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 Diablo Canyon 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 adequately 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 Diablo Canyon TS 5.5.16.

3.7 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals 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 (Reference 1).

In section 3.3.1, Methodology, of the enclosure to the LAR, the licensee states that the plant-specific risk assessment follows the guidelines in NEI 94-01, Revision 3-A and 2-A; the methodology described in EPRI TR-104285 (Reference 24), EPRI TR-1018243 (also identified as EPRI TR-1009325, Revision 2-A (Reference 9)); the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance from November 2001 (Reference 25); the NRC regulatory guidance on the use of PRA as stated in RG 1.200, Revision 3 (Reference 26) as applied to ILRT interval extensions; and the NRC regulatory guidance on risk insights in support of a request for a plants licensing basis as outlined in RG 1.174, Revision 3 (Reference 10).

Additionally, the licensee applied the methodology from Calvert Cliffs Nuclear Power Plant (Reference 27) to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during extended test interval.

RG 1.163, Revision 1, endorses EPRI TR-1009325 Revision 2-A, subject to the regulatory positions identified in section C of this RG. In table 3.3.1-1, RG 1.163, Revision 1, Section C, Staff Regulatory Guidance, of the enclosure to the LAR, the licensee addressed Regulatory Positions 7 through 10, which are consistent with the four conditions imposed by the NRC staff in section 4.2 of the NRC SE dated June 25, 2008 (Reference 28), for the use of EPRI Report 1009325, Revision 2. 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.

The licensee addresses Diablo Canyons PRA technical adequacy in the LAR enclosure, section 3.3.2, PRA Technical Adequacy. As discussed in section 3.3.2, the Diablo Canyon risk assessment performed to support the ILRT application utilized the current PRA model of record, DC05A Application Model, which the licensee completed in June 2023. The licensee explains its approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. In LAR section 3.3.2, under PRA Maintenance and Update, the licensee provides further details of the process used to ensure its PRA model reflects the as-built and as-operated plant.

In section 3.3.2 of the LAR enclosure, under PRA Self-Assessment and Peer Review, the licensee states that the internal events PRA model was subject to a full scope peer review in December 2012. Subsequently, all findings were resolved by either a PRA model revision or a documentation update. A formal facts and observations closure review was completed in July 2023 per NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard (Reference 29), using the NRC-accepted process documented in the NEI letter to the NRC Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-out of Facts and Observations (F&Os) (Appendix X process), dated February 21, 2017 (Reference 30) process, with a final report issued stating that all F&Os have been closed and meet Capability Category II or higher. Therefore, in accordance with RG 1.200, Revision 3, no F&Os associated with the internal PRA (including internal floods) were provided in the LAR. 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 as defined in NEI-07-12, Revision 1, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines (Reference 31), against the requirements of ANSI/ANS-58.23-2007, Fire PRA Methodology (Reference 32), in January 2008. A follow-on fire PRA peer review was conducted in December 2010 against the requirements of section 4 of the ASME/ANS RA-Sa-2009 PRA Standard, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008 (the ASME/ANS 2009 PRA standard) (Reference 33). In September 2018, an independent assessment of the F&Os was conducted using the Appendix X process. In conjunction with the closure review, a focused-scope peer review was performed for the fire PRA upgrade using the general process defined in NEI 07-12 and covered the relevant requirements in the ASME/ANS 2009 PRA Standard and the guidance in RG 1.200, Revision 2.

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 adequate to assess the changes to ILRT frequencies.

The licensees seismic PRA model was subject to a full scope review in January 2013 against the ASME/ANS 2009 PRA Standard and a second peer review using the updated 2013 version of the ASME/ANS PRA standard, ASME/ANS RA-Sb-2013, Addendum B to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 34), in June 2017. All F&Os were closed out by an independent assessment in February 2018 in accordance with the guidance of Appendix X process. Therefore, no F&Os associated with the fire PRA were provided in the LAR. The NRC staff concluded that the seismic PRA model was appropriately peer-reviewed, consistent with RG 1.200, Revision 3, and thus adequate to assess the changes to ILRT frequencies.

The licensee states that all other external hazards were screened out with the original external event analysis and updated analysis performed in 1988 and November 2016, respectively. The methods used were consistent with the screening and assessment process in the supporting requirements of Part 6 of the ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200, Revision 3.

Based on review of the above information, 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, 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 the licensees analysis associated with extending the Type A ILRT frequency:

1. Since the Diablo Canyon design does not rely on overpressure of containment to ensure adequate NPSH of the ECCS pumps, 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 1.12E-7/year using the EPRI guidance (this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included), with the baseline LERF as 2.71E-7/year. 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 7.36E-7/year using the EPRI guidance, and the baseline LERF is 7.55E-6/year. 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 0.040 person-rem/year. 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 0.906 percent. This estimated increase is below the acceptance guideline in accordance with EPRI Report No. 1009325, Revision 2-A 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, Regulatory Position 8 has been adequately 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 Diablo Canyon plant-specific risk assessment, based on the guidance provided in EPRI report No. 1009325, Revision 2-A.

Accordingly, Regulatory Position 9 has been adequately 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 [c]ontainment overpressure is not required in support of ECCS performance to mitigate design basis accidents at DCPP. Accordingly, Regulatory Position 10 has been satisfactorily addressed.

3.8 Technical Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRT and LLRT. The results of the recent ILRTs and of the LLRTs combined totals demonstrate acceptable performance and support a conclusion that the structural and leak-tight integrity of the primary containment 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 regulatory positions to demonstrate acceptability of adopting RG 1.163 Revision 1, and the limitations and conditions identified in NEI 94-01, Revision 3-A. 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 Diablo Canyon TS 5.5.16 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 third exception under Diablo Canyon TS 5.5.16 for the 15-year Type A test interval beginning May 4, 1994, for Unit 1 and April 30, 1993, for Unit 2. The exception was for activities that have already taken place and are no longer applicable. Therefore, the NRC staff finds that the deletion of third exception under Diablo Canyon TS 5.5.16 is acceptable because the deleted information is not required under 10 CFR 50.36(c)(5).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments on December 5, 2024. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facilitys 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 October 1, 2024 (89 FR 79970), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

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

6.0 CONCLUSION

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

7.0 REFERENCES

1.

Rogers, J. E., Pacific Gas and Electric Company, letter to U.S. Nuclear Regulatory Commission, License Amendment Request 24-03, Revision to Technical Specification 5.5.16 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies, dated July 31, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24213A331).

2.

Rogers, J. E., Pacific Gas and Electric Company, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information for License Amendment Request 24-03, Revision to Technical Specification 5.5.16 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies, dated February 12, 2025 (ML25043A418).

3.

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7.

Bloom, S. D., U.S. Nuclear Regulatory Commission, letter to Mr. G. M. Rueger, Pacific Gas and Electric Company, Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No. 1 (TAC No. M94379) and Unit No. 2 (TAC No. M94380), dated March 1, 1996 (ML022390548).

8.

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Report, 94-01, Revision 3, Industry Guideline for implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), dated June 8, 2012 (ML121030286).

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Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report No. 1009325, Revision 2-A, Palo Alto, CA, October 2008.

10.

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Principal Contributors: Brian Lee, NRR Clint Ashley, NRR Qin Pan, NRR Jerry Dozier, NRR Yi-Lun Chu, NRR Rob Atienza, NRR Ahsan Sallman, NRR Date: March 27, 2025

ML25063A335

• via e-mail OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA*

NRR/DSS/STSB/BC*

NRR/DSS/SCPB/BC*

NAME SLee PBlechman SMehta MValentin DATE 3/4/2025 3/12/2024 3/3/2025 3/3/2025 OFFICE NRR/DEX/ESEB/BC*

NRR/DRA/ARCB/BC*

NRR/DSS/SNSB/BC*

OGC*

NAME ITseng KHsueh DMurdock SGellen DATE 3/4/2025 12/22/2024 2/28/2025 3/27/2025 OFFICE NRR/DORL/LPL4/BC*

NRR/DORL/LPL4/PM*

NAME TNakanishi SLee DATE 3/27/2025 3/27/2025