ML25253A408

Issuance of Amendment Nos. 289 and 273 Changes to Technical Specifications for Leak Rate Testing

ML25253A408
Person / Time
Site:	Susquehanna
Issue date:	09/23/2025
From:	Audrey Klett Plant Licensing Branch 1
To:	Casulli E Susquehanna
Klett A
References
EPID L-2024-LLA-0148
Download: ML25253A408 (1)

Text

September 23, 2025 Mr. Edward Casulli Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard NUCSB3 Berwick, PA 18603-0467

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2ISSUANCE OF AMENDMENT NOS. 289 AND 273 RE: CHANGES TO TECHNICAL SPECIFICATIONS FOR LEAK RATE TESTING (EPID L-2024-LLA-0148)

Dear Mr. Casulli:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 289 and 273 to Renewed Facility Operating License Nos. NPF-14 and NPF-22 for the Susquehanna Steam Electric Station (Susquehanna), Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TS) in response to Susquehanna Nuclear, LLCs application dated November 1, 2024, as supplemented by letter dated June 26, 2025. These amendments revise TS 5.5.12, Primary Containment Leakage Rate Testing Program, in part, by adopting the guidance specified in NRC Regulatory Guide 1.163, Revision 1, Performance-Based Containment Leak-Test Program, which endorses, with conditions, the Nuclear Energy Institute (NEI) report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J. A copy of the related safety evaluation is also enclosed. A notice of issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Audrey Klett, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosures:

1. Amendment No. 289 to License No. NPF-14

2. Amendment No. 273 to License No. NPF-22

3. Safety Evaluation cc: Listserv SUSQUEHANNA NUCLEAR, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-387 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 289 Renewed License No. NPF-14

1.

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has found that:

A.

The application for the amendment filed by Susquehanna Nuclear, LLC, dated November 1, 2024, as supplemented by letter dated June 26, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the 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 set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the technical specifications and renewed facility operating license, as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-14 is hereby amended to read, in part, as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 289, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 23, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.09.23 12:47:31 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 289 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following page of the renewed facility operating license with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT Page 3 Page 3 Replace the following pages of the Appendix A technical specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 5.0-18 5.0-18 Renewed Operating License No. NPF-14 Amendment No. 289 (3)

Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, posses, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, posses, 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)

Susquehanna Nuclear, LLC, 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 nor or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Susquehanna Nuclear, LLC is authorized to operate the facility at reactor core power levels not in excess of 3952 megawatts thermal in accordance with the conditions specified herein. The preoperational tests, startup tests and other items identified in License Conditions 2.C.(36), 2.C.(37), 2.C.(38), and 2.C.(39) to this license shall be completed as specified.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 289, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

For Surveillance Requirements (SRs) that are new in Amendment 178 to Facility Operating License No. NPF-14, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 178. For SRs that existed prior to Amendment 178, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 178.

Programs and Manuals 5.5 (continued)

SUSQUEHANNA - UNIT 1 5.0-18 Amendment 178, 202, 209, 241, 246, 289 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.12 Primary Containment Leakage Rate Testing Program A program shall be established, implemented, and maintained to comply with 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, as modified by the following exceptions:

a. 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.

b. 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.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48.6 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 1% of the primary containment air weight per day.

SUSQUEHANNA NUCLEAR, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-388 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 273 Renewed License No. NPF-22

1.

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has found that:

A.

The application for the amendment filed by Susquehanna Nuclear, LLC, dated November 1, 2024, as supplemented by letter dated June 26, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the 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 and renewed facility operating license, as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-22 is hereby amended to read, in part, as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 23, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.09.23 12:48:09 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 273 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following page of the renewed facility operating license with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT Page 3 Page 3 Replace the following page of the Appendix A technical specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 5.0-18 5.0-18 Renewed Operating License No. NPF-22 Amendment No. 273 (3)

Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, posses, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, posses, 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)

Susquehanna Nuclear, LLC, 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 nor or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Susquehanna Nuclear, LLC is authorized to operate the facility at reactor core power levels not in excess of 3952 megawatts thermal in accordance with the conditions specified herein. The preoperational tests, startup tests and other items identified in License Conditions 2.C.(20), 2.C.(21), 2.C.(22), and 2.C.(23) to this license shall be completed as specified.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

For Surveillance Requirements (SRs) that are new in Amendment 151 to Facility Operating License No. NPF-22, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 151. For SRs that existed prior to Amendment 151, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 151.

Programs and Manuals 5.5 (continued)

SUSQUEHANNA - UNIT 2 5.0-18 Amendment 151, 176, 183, 219, 224, 273 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.12 Primary Containment Leakage Rate Testing Program A program shall be established, implemented, and maintained to comply with 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, as modified by the following exceptions:

a.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10CFR 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.

b.

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.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48.6 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 1% of the primary containment air weight per day.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR AMENDMENT NO. 289 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-14 AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-22 SUSQUEHANNA NUCLEAR, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 DOCKET NOS. 50-387 AND 50-388

1.0 INTRODUCTION

1.1 Overview By application dated November 1, 2024 [1], as supplemented by letter dated June 26, 2025 [2],

Susquehanna Nuclear, LLC (the licensee) requested changes to the technical specifications (TS) for Susquehanna Steam Electric Station, Units 1 and 2 (Susquehanna, Units 1 and 2). The proposed changes would revise the TS for the primary containment leakage rate testing program by extending testing intervals and frequencies, replacing references, and deleting obsolete information.

The supplement dated June 26, 2025 [2], 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 January 21, 2025 [3].

1.2 Proposed Changes In its license amendment request (LAR) [1], as supplemented [2], the licensee proposed the following changes:

increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Regulatory Guide (RG) 1.163, Revision 1, Performance-Based Containment Leak-Test Program [4],1 which endorses, with conditions, Nuclear Energy Institute (NEI) report NEI 94-01, Revision 3-A, Industry 1 References to RG 1.163 in this safety evaluation are to Revision 1 of this guide unless another revision is explicitly cited.

Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J [5]2 adopt an extension of the containment isolation valve leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR Part 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 [5]

adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2020, Containment System Leakage Testing Requirements [6]

adopt a more conservative allowable test interval extension of nine months, for Types A, B, and C leakage rate tests in accordance with NEI 94-01 [5]

extend each units drywell-to-suppression chamber bypass leak rate test (DWBT) TS Surveillance Requirement (SR) 3.6.1.1.2 frequency from 120 months (10 years) to 180 months (15 years)

The licensee proposed to revise each units TS 5.5.12, Primary Containment Leakage Rate Testing Program, by replacing the reference to RG 1.163, dated September 1995 [7], with a reference to RG 1.163, Revision 1, dated June 2023 [4]. This document would be used to implement the performance-based leakage testing program in accordance with 10 CFR Part 50, Appendix J, Option B.

The licensee also proposed deleting exception c from each units TS 5.5.12 because it is no longer applicable. TS 5.5.12, exception c, identifies the performance dates of Type A tests approved for deferral via Amendment Nos. 202 and 176 [8] for Susquehanna, Units 1 and 2, respectively.

1.3 Description of Containment Leakage Testing Section 3.1.1, Description of the Containment, of attachment 1 to the LAR [1] describes the boiling-water reactor, Mark II primary containments for Susquehanna, Units 1 and 2, which are enclosures for the reactor vessels, the reactor coolant recirculation loops, and other connections of the reactor coolant systems.

The NRC requires containment leakage tests for verifying the leak-tight integrity of the primary reactor containment and systems and components that penetrate containment. There are three types of containment leakage tests:

Type A tests, which measure the primary reactor containment overall integrated leakage rate after the containment has been completed and is ready for operation, and at periodic intervals thereafter 2 References to NEI 94-01 in this safety evaluation are to Revision 3-A of the report unless another revision is explicitly cited.

Type B tests, which detect local leaks and measure leakage across each pressure-containing or leakage-limiting boundary for specified primary reactor containment penetrations, such as seals, gaskets, and expansion bellows Type C tests, which measure containment isolation valve leakage rates Type A tests are referred to as ILRTs, and Type B and C tests are referred to as local leakage rate tests (LLRTs).

2.0 REGULATORY EVALUATION

2.1 Requirements The regulations in 10 CFR 50.54(o) require that the primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in 10 CFR Part 50, Appendix J. Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components that penetrate the containment. Appendix J also 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 testing requirements in Appendix J ensure that:

(a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TS, and (b) integrity of the containment structure is maintained during the service life of the containment.

Susquehanna is required to comply with Option B, as prescribed in TS 5.5.12. The Option B performance-based containment leakage rate testing provisions for Type A, B, and C testing do not alter the basic method by which Appendix J leakage rate testing is performed; however, they do alter the frequency at which Type A, B, and C containment leakage tests must be performed.

Under this performance-based option, the leakage testing frequency is based on an evaluation of the as-found leakage history, which provides assurance that leakage limits will be maintained.

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

The regulations in 10 CFR 50.36(c), Technical specifications, paragraph (3) state that the TS must include SR, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. TS 3.6.1.1, Primary Containment, SR 3.6.1.1.2 requires the licensee to verify the DWBT leakage value at a frequency in accordance with the primary containment leakage rate testing program and 24 months, only when required after two consecutive test failures and continues until two consecutive passed tests.

The regulations in 10 CFR 50.36(c)(5) state that the TS must include administrative controls, which 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.

TS 5.5.12 is in the administrative controls section of the TS. TS 5.5.12 prescribes requirements for the primary containment leakage rate testing program, maximum allowable primary containment leakage rate, and leakage rate acceptance criteria.

2.2 Guidance NEI 94-01, Revision 3-A [5] 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 LLRT intervals beyond 60 months. On June 8, 2012, the NRC staff published a safety evaluation [9] for NEI 94-01, Revision 3 [10] in which the staff concluded that, subject to limitations and conditions, 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. The staffs safety evaluation was incorporated into NEI 94-01, Revision 3 and subsequently issued as NEI 94-01, Revision 3-A.

The Electric Power Research Institute (EPRI) technical report 1009325 (EPRI-1009325),

Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [11],3 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. The NRC staff endorsed EPRI-1009325 with limitations and conditions in its safety evaluation dated June 28, 2008 [12]. 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 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 [13],4 and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications [14], to support changes to surveillance test intervals.

RG 1.163 [4] endorses NEI 94-01 [5] for implementing Option B of Appendix J to 10 CFR Part 50, subject to the regulatory positions listed in section C of RG 1.163. The guidance in NEI 94-01 [5] includes extending Type A test intervals up to 15 years and Type C test intervals up to 75 months. RG 1.163 also endorses EPRI-1009325 [11], subject to the applicable regulatory positions listed in section C of the guide. In addition, RG 1.163 also endorses ANSI/ANS 56.8-2020 [6] regarding acceptable industry standards for technical methods and techniques for performing Type A, B, and C tests.

3.0 TECHNICAL EVALUATION

Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The staff evaluated the request to determine whether the 3 References to EPRI-1009325 in this safety evaluation are to Revision 2-A of this report unless another revision is explicitly cited.

4 References to RG 1.174 in this safety evaluation are to Revision 3 of this guide unless another revision is explicitly cited.

proposed changes are consistent with the regulations, licensing basis, and guidance discussed in section 2 of this safety evaluation. The staff reviewed the proposed changes to TS to determine whether they meet the requirements of Appendix J to 10 CFR Part 50, 10 CFR 50.55a, and 10 CFR 50.36 and provide reasonable assurance that operation with the proposed changes will not endanger the health and safety of the public.

3.1 Evaluation of ILRT and LLRT (Types A, B, and C Testing) History 3.1.1 ILRT (Type A Testing) History In Section 3.2.4 of Attachment 1 to the LAR [1], the licensee provided the Susquehanna units ILRT history. The licensee provided the test result summaries in Table 3.2.4-1, Periodic Type A ILRT Results, and Table 3.2.5-1, [Susquehanna] ILRT Test Results Verification of Current Extended ILRT Interval, of Attachment 1 to the LAR. There have been six ILRTs performed on each unit to date. For a Unit 2 test result that was unsatisfactory in 1986, the licensee identified the cause of the unsatisfactory test, and subsequent tests were satisfactory. The NRC staff reviewed the past ILRT results and finds that there has been a substantial margin maintained relative to the performance criterion in TS 5.5.12. Because the last two Type A tests for each unit had as found test results well within the current maximum allowable containment leakage rate specified in TS 5.5.12, the NRC staff finds that extending the Type A test frequency to 15 years is consistent with NEI 94-01 [5] and the regulatory positions in RG 1.163 [4]. Based on its review, the NRC staff concludes that the ILRT test results provide reasonable assurance that the licensee will maintain containment overall leakage below the design-basis leak rate, consistent with the requirements in TS 5.5.12 and 10 CFR Part 50, Appendix J, Option B, when using the proposed test frequency of 15 years.

3.1.2 Primary Containment Leakage Rate Testing (Types B and C Testing) Program In Section 3.4.6 of Attachment 1 to the LAR [1], the licensee provided the results of its primary containment leakage rate Type B and Type C testing program, which include leakage rates, component failures, and corrective actions. Tables 3.4.6-1 and 3.4.6-2, Unit 1 [and 2, respectively] Type B and Type C LLRT Combined As-Found/As-Left Trend Summary, of to the LAR summarize the LLRT as-found and as-left trends for the five previous LLRT tests for each unit (for a total of 10 refueling and inspection outages). The NRC staff confirmed that the as-found minimum, as-left minimum, and as-left maximum pathway leakage rate test results satisfy the acceptance criteria in Section 8.0 of NEI 94-01 [5].

Tables 3.4.6-3 and 3.4.6-4, Unit 1 [and 2, respectively] Type B and Type C LLRT Program Implementation Review, of Attachment 1 to the LAR [1] summarize component performance and corrective actions during the previous two outages for both units. Based on its review of these tables, the NRC staff has reasonable assurance that the licensees Type B and Type C testing and corrective action programs provide for identifying and correcting issues with valves and penetrations.

Based on its review, the NRC staff has reasonable assurance that the licensee is effectively implementing its Type B and Type C leakage rate test program, as required by Option B of 10 CFR Part 50, Appendix J. Therefore, the staff finds that extending the containment isolation valve leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR Part 50, Appendix J, Option B to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with the guidance in NEI 94-01 [5], is acceptable.

3.1.3 DWBT Test History By letter dated September 6, 1996 [15], the NRC issued Amendment Nos. 160 and 131 for Susquehanna, Units 1 and 2, respectively, approving the use of 10 CFR Part 50, Appendix J, Option B, along with the corresponding frequency change to the DWBT. As a result of those amendments, TS SR 3.6.1.1.2 was revised to include wording to conduct the DWBT to coincide with the Type A test.

In Section 3.2.6, Bypass Leak Rate Test Risk Assessment, of Attachment 1 to the LAR [1], the licensee describes the DWBT history. Table 3.2.6-1, [Susquehanna] Drywell-to-Suppression Chamber Bypass Leakage Testing, of Attachment 1 to the LAR lists results for the last two DWBTs at Susquehanna, Units 1 and 2, which did not include any failures. The licensee evaluated the history of test results and stated in its LAR that the leakage is less than three percent of the acceptance criteria, which is set at an order of magnitude below the design-basis limit.

The NRC staff reviewed the past DWBT results presented in the LAR [1] and found that there was substantial margin maintained relative to the performance criterion in TS SR 3.6.1.1.2.

Therefore, the NRC staff concludes that the past DWBT test results provide reasonable assurance that the drywell bypass leakage will be maintained below the design-basis value, consistent with the requirements of TS SR 3.6.1.1 and 10 CFR Part 50, Appendix J, Option B, when using the proposed test frequency of 15 years.

Based on the above evaluation, the staff finds that the licensee would continue to meet 10 CFR 50.36(c)(3) with the proposed changes because the SR assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.1.4 Type B and Type C Components on Extended Leakage Test Intervals In Section 3.4.7 of Attachment 1 to its LAR [1], as corrected by its supplement [2], the licensee provided the number of Type B and C tested components, the number of those components eligible for an extended leakage test interval, and the number of those eligible components that are on an extended interval. To develop this information, the licensee reviewed the population of LLRT components for each reactor unit to identify the Type B and Type C components eligible for extended interval and then reviewed the eligible population performance history to identify those on an extended interval. Per the LAR, as supplemented, the Type B and Type C components eligible for an extended leakage test interval may change because of performance data and reevaluation of eligible components based on actual testing requirements.

As of the date of the supplement [2], for Unit 1, there are 51 Type B tested components, 43 of which are eligible for an extended interval, and 43 of the eligible components are on an extended interval. For Unit 1, there are 199 Type C tested components, 142 of which are eligible for an extended interval, and 134 components are on an extended interval. For Unit 2, there are 51 Type B tested components, 43 of which are eligible for an extended interval, and 43 of the eligible components are on an extended interval. For Unit 2, there are 201 Type C tested components, 148 of which are eligible for an extended interval, and 141 of the eligible components are on an extended interval.

In its LAR [1], the licensee requested to modify TS 5.5.12 to allow, in part, adopting an extension of the Type C leakage rate testing interval from the 60 months currently permitted by 10 CFR Part 50, Appendix J, Option B, to 75 months for Type C leakage rate testing of selected components, in accordance with NEI 94-01 [5], and adopting a more conservative allowable test interval extension of 9 months, for Type A, B and C leakage rate tests in accordance with NEI 94-01. The NRC staff reviewed the information provided in the LAR and its supplement [2]

describing the licensees implementation of the provisions in NEI 94-01, as accepted by the NRC [9], to identify the Type B and C components that are eligible for extended leakage rate testing intervals. Based on its review, the NRC staff determined that the licensees description of its evaluation of the performance of Type B and C components summarized in the LAR, as supplemented, provides reasonable assurance that the licensee will properly implement the provisions of NEI 94-01, as accepted by the NRC, when determining the Type B and C components that are eligible for an extended leakage test interval and condensing the interval when required by the performance of those components.

In Section 3.2.1, Chronology of Testing Requirements of 10 CFR Part 50, Appendix J, of to the LAR [1], the licensee described leakage testing requirements related to containment, including the Type C tests to measure containment isolation valve leakage rates.

The NRC regulations require the licensee to implement the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code, as incorporated by reference in 10 CFR 50.55a for the inservice testing program to assess the operational readiness of specific pumps, valves, and dynamic restraints. The licensee did not indicate any proposed changes to its inservice testing program from the implementation of the proposed changes to TS 5.5.12 in its LAR. If any inservice testing program changes are planned that would not meet the NRC regulatory requirements in the ASME OM Code as incorporated by reference in 10 CFR 50.55a, then the licensee is required to submit requests for those changes in accordance with NRC regulations.

Based on the information provided in the LAR [1], as supplemented [2], the NRC staff finds that the licensee has adequately supported the proposed extension of the leakage test intervals for Type B and C components.

3.2 Containment Inspection Evaluation 3.2.1 Service Level I Coating Systems In Section 3.4.1, Monitoring the Performance of Service Level 1 Coatings Systems, of to the LAR [1], the licensee described the protective coatings monitoring program for coatings inside the primary containment that could adversely affect systems, structures, or components important to safety. The licensee also provided each units quantities of qualified and unqualified coatings on the drywell and suppression pool. The LAR indicated that the assessment of the containment coatings is based on the American Society for Testing and Materials (ASTM) standard, ASTM D5163-08, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants [16]. In Section 3.4.2, Recent Service Level I Coatings Inspection Reports, of to the LAR, the licensee provided data sheets of recent coatings inspections performed between 2018 and 2023 during refueling and inspection outages. The NRC staff reviewed the data sheets and finds that based on the licensees findings of the inspected coatings and corrective actions, there is reasonable assurance that the licensees protective coatings monitoring program will identify and address any issues with the coatings.

3.2.2 Containment Inservice Inspection (IWE/IWL) Program Plan On March 19, 2007, the NRC issued Amendment Nos. 241 and 219 [17] for Susquehanna, Units 1 and 2, respectively, which allowed the visual examination of the containment pursuant to ASME Boiler and Pressure Vessel (BPV) Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Subsections IWE and IWL to fulfill the requirements of RG 1.163, Revision 0 [7]. In Section 3.4.3, [Susquehanna] Units 1 and 2, 3rd Interval Containment Inservice Inspection (IWE/IWL) Program Plan, of Attachment 1 to the LAR [1], the licensee provided information related to the containment inservice inspection (CISI) program.

The licensee stated that the CISI program is established in accordance with 10 CFR 50.55a, complies with ASME BPV Code,Section XI, and implements the requirements of the Updated Final Safety Analysis Report (UFSAR) [18], Section 5.2.4, Inservice Inspection and Testing of Reactor Coolant Pressure Boundary, and Section 6.6, Inservice Inspection of Class 2 and 3 Components. The fourth ten-year interval ISI and third ten-year interval CISI program plans are based on the requirements of the ASME BPV Code,Section XI, 2007 Edition with the 2008 Addenda with conditions defined in 10 CFR 50.55a(b)(2).

In Section 3.4.5, Results of Recent Containment Examinations, of Attachment 1 to the LAR [1],

the licensee presented the results of recent visual inspections per ASME BPV Code,Section XI, Subsections IWE and IWL examinations conducted during refueling and inspection outage for Units 1 and 2 between 2016 and 2023. Tables 3.4.5-1 through 3.4.5-8 in Attachment 1 to the LAR summarized the IWE and IWL examinations and showed that the results of the most recent IWE examinations have been acceptable except for a few locations of pitting at the bottom of the suppression pool; however, none of the pitting depths approached the acceptance criteria limit prescribed by the ASME BPV Code,Section XI, Subsection IWE. Tables 3.4.5-7 and 3.4.5-8 in Attachment 1 to the LAR showed that the results of the most recent IWL examinations for both Units 1 and 2 are acceptable. Based on review of the program descriptions and results of recent examinations, the NRC staff finds there is reasonable assurance that the licensees CISI program will continue to manage the aging effects such that the containment will continue to perform its safety-related function.

3.2.3 Operating Experience In Section 3.5, Operating Experience of Attachment 1 to the LAR [1], the licensee listed and evaluated site-specific and industry events for their impact on the containment. The licensee provided the results of the review of the applicable operating experience, NRC generic communications, and NRC inspection reports and described how it uses such information to inform the programs for maintaining the overall containment integrity. The NRC staff reviewed Sections 3.5.1 through 3.5.6 of Attachment 1 to the LAR and the related UFSAR [18] sections and drawings for applicability and consistency. The NRC staff determined that the licensee adequately addressed the relevant regulatory requirements, guidance, and operating experience listed in Section 3.5 of Attachment 1 to the LAR through inspection and aging management programs. Therefore, the staff finds that the licensees CISI program provides reasonable assurance that the containment will maintain its capability to perform its safety-related function.

3.2.4 License Renewal Aging Management In Section 3.6, License Renewal Aging Management, of Attachment 1 to the LAR [1], the licensee described its aging management programs for containment leakage rate testing, ISI Program - IWE, and ISI Program - IWL, which are also part of the supporting basis for the LAR. The NRC staff reviewed the program descriptions and, based on these descriptions and the various test and inspection results provided in the LAR, the NRC staff determined that there is reasonable assurance that the programs will continue to manage aging effects on the concrete containment, steel liner, and containment penetrations such that the applicable structures and components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

3.3 Evaluation of RG 1.163, Revision 1, Section C Regulatory Positions Responses RG 1.163 [4] endorses NEI 94-01 [5], subject to the regulatory positions identified in section C of this RG. In Table 3.7.1-1, RG 1.163 Revision 1, Staff Regulatory Guidance, in Attachment 1 to the LAR [1], the licensee provided a response to each of these regulatory positions. The staffs evaluation of the licensees responses to regulatory positions 1 through 6 is below.

RG 1.163 also includes regulatory positions 7 through 10 related to the risk assessment methodology. The NRC staffs evaluation of the licensees responses to those positions are in section 3.5 of this safety evaluation. Based on the evaluation in sections 3.3.1 through 3.3.6 and section 3.5.1 below, the NRC staff finds it acceptable for the licensee to adopt RG 1.163, Revision 1 [4] as the implementation document listed in TS 5.5.12.

3.3.1 Regulatory Position 1 Regulatory position 1 of RG 1.163 [4] states, in part:

... ANSI/ANS-56.8-2020 as approved by this RG may be used in lieu of ANSI/ANS 56.8-2002 without a LAR if (1) the licensees TS incorporate NEI 94-01, Revision 3-A and (2) the licensees TS do not explicitly reference the 2002 ANSI/ANS standard and there is no other license provision that would necessitate a LAR to use ANSI/ANS-56.8-2020.... [F]or calculating the Type A leakage rate, the licensee should use the performance leakage rate definition in NEI 94-01, Revision 3-A, in lieu of that in ANSI/ANS-56.8-2002 or ANSI/ANS-56.8-2020.

In Table 3.7.1-1 of Attachment 1 to the LAR [1], the licensees response to regulatory position 1 states that it will use ANSI/ANS 56.8-2020 and the definition in Section 5.0 of NEI 94-01 [5]. The NRC staff finds that the licensees response is consistent with the RG [4]. Therefore, the NRC staff concludes that the licensee adequately addressed regulatory position 1.

3.3.2 Regulatory Position 2 Regulatory position 2 of RG 1.163 [4] 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 a Type A test interval extension. In Table 3.7.1-1 of Attachment 1 to the LAR [1],

the licensees response to regulatory position 2 refers to Section 3.4.3 (Tables 3.4.3-3 and 3.4.3-4) of Attachment 1 to the LAR.

In Table 3.4.3-1 through Table 3.4.3-4 in Attachment 1 to the LAR [1], the licensee provided a schedule for the third and fourth CISI intervals for both units to perform examinations required by ASME BPV Code,Section XI, Subsections IWE and IWL. Visual examinations of metal components are required for Category E-A, E-C, and E-G components during the 10-year CISI interval. The schedule is detailed and organized by planned outage and period within the 10-year interval. General visual examinations of Category L-A concrete components are conducted every five years as part of Subsection IWL-2410 to the ASME BPV Code,Section XI requirement. Furthermore, Section 3.6.1, Containment Leakage Rate Test Program [Aging Management Program], of Attachment 1 to the LAR states, A general visual inspection of the accessible interior and exterior surface of the primary containment structures and components is performed prior to Type A test. The Susquehanna units are currently in their fourth CISI interval, which began on June 1, 2024, and is scheduled to conclude on May 31, 2034. The NRC staff finds that the schedule described in Tables 3.4.3-1 through 3.4.3-4 and Section 3.6.1 of Attachment 1 to the LAR meets regulatory position 2.

3.3.3 Regulatory Position 3 Regulatory position 3 of RG 1.163 [4] states, in part, The LAR should address the areas of the containment structure potentially subject to degradation. In Table 3.7.1-1 of Attachment 1 to the LAR [1], the licensees response to regulatory position 3 refers to Section 3.4.3 of Attachment 1 of the LAR.

As described in Section 3.4.3, [Susquehanna] Units 1 and 2, 3rd Interval Containment Inservice Inspection (IWE/IWL) Program Plan, of Attachment 1 to the LAR [1], the licensees CISI program is developed in accordance with the requirements of ASME BPV Code,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a, Codes and standards. The current fourth 10-year CISI program plans are based on the 2007 edition of ASME BPV Code,Section XI with 2008 Addenda.

The NRC staff reviewed the ISI program IWE and IWL aging management programs described in described in Sections 3.6.2 and 3.6.3 of Attachment 1 to the LAR [1], respectively, and finds that the licensee provided an acceptable level of information to demonstrate the accessible areas of containment structures potentially subject to degradation are effectively managed. The NRC staff reviewed how the licensee evaluated inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation of inaccessible areas. In its supplement [2], the licensee clarified that the ISI program plan requires that, for each inaccessible area identified for evaluation, the information specified in 10 CFR 50.55a(b)(2)(viii)(E) and 10 CFR 50.55a(b)(2)(ix)(A) is provided in the ISI summary report. Furthermore, the licensee indicated that results of past IWE and IWL inspections have not identified conditions in accessible areas that could indicate the presence of or result in in degradation of inaccessible areas and, therefore, evaluations of inaccessible areas per the regulatory requirements have not been necessary. Based on this information, the NRC staff finds that the licensees response adequately addressed and is consistent with regulatory position 3.

3.3.4 Regulatory Position 4 Regulatory position 4 of RG 1.163 [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 of Attachment 1 to the LAR [1], the licensees response to regulatory position 4 states that there are no major modifications planned. Based on this information, the NRC staff finds that regulatory position 4 is not applicable.

3.3.5 Regulatory Position 5 Regulatory position 5 of RG 1.163 [4], 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...).

In Table 3.7.1-1 of Attachment 1 to the LAR [1], the licensees response to regulatory position 5 states that the licensee will follow the requirements of Section 9.1 of NEI 94-01 [5] and demonstrate to the NRC staff that an unforeseen emergent condition exists if an extension beyond the 15-year interval is required. The NRC staff finds that the licensees response adequately addressed and is consistent with regulatory position 5.

3.3.6 Regulatory Position 6 Regulatory position 6 of RG 1.163 [4] addresses new reactor plants licensed under 10 CFR Part 52 and, therefore, is not applicable to Susquehanna.

3.4 Evaluation of NEI 94-01 Conditions of Use As discussed in section 2.2 of this safety evaluation, the NRC endorsed NEI 94-01 [5], subject to two conditions. In its LAR [1], as supplemented [2], the licensee proposed to use RG 1.163 [4], which endorses NEI 94-01 [5] as the implementation document for the leak rate testing program. Accordingly, the licensee requested to adopt the testing criteria of ANSI/ANS 56.8-2020 [6] as part of its licensing basis. As discussed in section 2.0, Purpose and Scope, of NEI 94-01 [5], where technical guidance overlaps between NEI 94-01 and the ANSI standard, the guidance in NEI 94-01 takes precedence.

In Section 3.7.2, Limitations and Conditions Applicable to NEI 94-01, Revision 3-A, of to the LAR [1], the licensee describes how it proposes to meet the two conditions on the use of NEI 94-01 [5]. The NRC staff reviewed the licensees responses in Section 3.7.2 of Attachment 1 to the LAR for the three issues of condition 1 and the two issues of condition 2 of NRCs safety evaluation [9] for NEI 94-01 and finds that the licensees responses satisfactorily address each issue. Therefore, the staff determined that the licensee adequately addressed the conditions in section 4.0 of NRCs safety evaluation for NEI 94-01. Therefore, the NRC staff finds it acceptable for the licensee to adopt and incorporate RG 1.163 [4], which endorses NEI 94-01, as the implementation document in the proposed change to TS 5.5.16.

3.5 Evaluation of Risk Assessment In Enclosure 1, Risk Impact Assessment of Extending Susquehanna ILRT/DWBT Interval, and in Section 3.3, Plant Specific Confirmatory Analysis, of Attachment 1 to the LAR [1], the licensee provided a plant-specific risk assessment for permanently extending the currently allowed containment Type A ILRT and DWBT interval. In Section 3.3.1, Methodology, of to the LAR, the licensee discussed the guidelines used to perform the assessment. The licensee addressed each of the four conditions on the use of EPRI-1009325 [11], which are listed in section 4.2 of the NRC safety evaluation [12], and conditions 7 through 10 of RG 1.163 [4]. The NRC staffs evaluation of the LAR against each condition is provided in section 3.5.1 below.

3.5.1 Plant-Specific Risk Evaluation 3.5.1.1 PRA Acceptability The NRCs safety evaluation [12] (for EPRI-1009325 [11]) and RG 1.163 [4] stipulate that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the guidance in RG 1.200, Probabilistic Risk Assessment Results for Risk-Informed Activities, relevant to the ILRT extension application.5 In section C of RG 1.163 [4], the NRC staff states that Capability Category I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension applications because approximate values of core damage frequency (CDF), large early release frequency (LERF), and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

In Section 3.3.2, PRA Technical Acceptability, of Attachment 1 and in Appendix A, PRA Technical Acceptability, of Enclosure 1 of the LAR [1], the licensee described its PRA technical acceptability. The risk assessment performed to support the ILRT and DWBT application utilized the current internal events, internal flooding, and fire PRA models of record. The licensee explained its approach to establishing and maintaining the technical acceptability and plant fidelity of the PRA models. This approach includes a proceduralized PRA maintenance and update process and the use of self-assessments.

In Section 3.3.2 of Attachment 1 to the LAR [1], the licensee described the peer reviews, independent and self-assessments, and closure of facts and observations for Susquehannas internal events, internal flooding, and internal fire PRAs. The NRC previously concluded [19]

that these PRAs were acceptable for use in the risk-informed completion time program. With respect to external events, RG 1.174 [13] indicates that (1) established acceptance guidelines are intended for comparison with a full-scope assessment of the change in the applicable risk metrics, (2) many PRAs are not full scope, and (3) PRA information of less than full scope may be acceptable. The methodology described in EPRI-1009325, which the NRC found satisfies the key principles of risk-informed decision-making of RG 1.174, explains that if the external event analysis is not of sufficient quality or detail to allow direct application of the methodology, then the quality or detail will be increased, or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order-of-magnitude estimate for contribution of the external event to the impact of the changed interval.

In Section 3.3.2 of Attachment 1 to the LAR [1], the licensee indicated that, based on the above information, it performed an analysis of the potential impacts from external events in a bounding fashion. This analysis references the currently available information for external events models and information to develop the bounding CDF associated with intact containment sequences.

5 The LAR references Revision 3 of RG 1.200 [22]. 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.

Specifically, as noted in Section 5.7 of Enclosure 1 to the LAR, the licensee utilizes the baseline seismic LERF and CDF, as calculated for use in the risk-informed completion time program [19],

and the probability of a class 3b event to estimate the frequency of class 3b events and the resulting change in LERF associated with an ILRT extension for both fire and seismic contributions to risk. The NRC staff finds that this method of estimating intact containment sequences, the frequency of class 3b events, and the change in LERF to be a conservative risk estimate and in accordance with the methodology in the EPRI Report No. 104285, A Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals [20], as updated by EPRI-1009325 [11].

Based on its review of the above information, the NRC staff finds that the licensee has adequately addressed the relevant findings and gaps from the peer reviews. Therefore, the NRC staff concludes that the internal events PRA model and external event contribution estimations used by the licensee are of sufficient quality to support the evaluation of changes to ILRT frequencies. Accordingly, the NRC staff determined that the licensee meets condition 7 of RG 1.163 [4], and condition 1 in section 4.2 of the NRC safety evaluation [12] for EPRI-1009325 [11].

3.5.1.2 Estimated Risk Increase The NRCs safety evaluation [12] (for EPRI-1009325 [11]) and RG 1.163 [4] stipulate 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 [13] and the clarification provided in the NRC safety evaluation for EPRI-1009325.

The NRCs safety evaluation for EPRI-1009325 defines a small increase in population dose as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, the NRCs safety evaluation for EPRI-1009325 defines a small increase in conditional containment failure probability (CCFP) as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests, which would require that the increase in CCFP be less than or equal to 1.5 percentage points. In Section 3.3 of Attachment 1 and in Section 7.0 of Enclosure 1 to the LAR, the licensee reported the results of the plant-specific risk assessment and its conclusions drawn from its analysis associated with extending the Type A ILRT frequency. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years to one test in 15 years.

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 [13], 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 among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.

Accordingly, the NRC staff determined that the licensee meets condition 8 of RG 1.163 [4], and condition 2 in section 4.2 of the NRC safety evaluation [12] for EPRI-1009325 [11].

3.5.1.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The NRCs safety evaluation [12] (for EPRI-1009325 [11]) and RG 1.163 [4] stipulate that for the methodology in EPRI-1009325 to be acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b in the EPRI methodology) shall be 100 La (percentage weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) instead of 35 La. As discussed in Section 3.0 of to the LAR [1], the licensee incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case. Therefore, the NRC staff concludes that the licensee used the 100 La value in the plant-specific risk assessment.

Accordingly, the NRC staff determined that the licensee meets condition 9 of RG 1.163 [4], and condition 3 in section 4.2 of the NRC safety evaluation [12] for EPRI-1009325 [11].

3.5.1.4 Containment Overpressure for Emergency Core Cooling System Performance The NRCs safety evaluation [12] (for EPRI-1009325 [11]) and RG 1.163 [4] stipulate that in instances where containment overpressure is relied upon for emergency core cooling system performance, a LAR is required to be submitted. In Table 3.3.1-1 of the Enclosure 1 to the LAR [1], the licensee stated that Susquehanna does not rely upon containment overpressure for emergency core cooling system performance. Accordingly, the NRC staff determined that condition 10 of RG 1.163 [4], and condition 4 in section 4.2 of the NRC safety evaluation [12] for EPRI-1009325 [11] are not applicable.

3.5.2 DWBT Risk Assessment Section 3.2.6 of Attachment 1 and Appendix B of Enclosure 1 to the LAR [1] describe the pressure suppression containment design, the design value for leakage area, and TS requirements for the maximum allowable bypass leakage flow area. Section 3.2.6 of and Appendix B.2 of Enclosure 1 to the LAR also describe the historical test results for the DWBTs, which did not identify any failures. The licensee stated in its LAR that the history of test results indicates that the typical leakage is less than three percent of the acceptance criterion, which is set at an order of magnitude below the design-basis limit.

Appendix B.4 of Enclosure 1 to the LAR summarizes the licensees deterministic thermal hydraulic analyses and results that the licensee performed as part of the DWBT risk assessment to identify the impact of increased drywell-to-suppression chamber leakage on the risk spectrum. Appendix B.5 of Enclosure 1 to the LAR listed the licensees findings based on the results of the deterministic studies and their PRA implications.

The NRC staff confirmed that the changes in CDF and LERF identified in the studies results meet the RG 1.174 [13] acceptance guidelines for very small risk change. The change in population dose rate is well below the acceptance criteria of less than or equal to 1.0 person-rem per year or less than 1 percent person-rem per year defined in the EPRI guidance. The change in CCFP of 0.014 percent is significantly below the EPRI guidance document acceptance criteria of less than 1.5 percent. As such, the NRC staff has reviewed the risk impact of the drywell-to-suppression chamber bypass leak rate test frequency extension provided by the licensee and concludes that the risk impact is acceptable for this application.

3.6 Evaluation of Technical Specifications Changes and Technical Evaluation Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff has reasonable assurance that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRT and LLRT and that the structural and leak-tight integrity of the primary containment is managed. The results of the recent ILRTs and of the LLRTs combined totals provide reasonable assurance of that the structural and leak-tight integrity of the primary containment 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 in RG 1.163 [4] and the limitations and conditions identified in NEI 94-01 [5]. 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 replacing the reference to RG 1.163, Revision 0 [7] with RG 1.163, Revision 1 [4] in TS 5.5.12 is acceptable and continues to meet 10 CFR 50.36(c)(5), and that extending the frequency of TS SR 3.6.1.1.2 for the drywell-to-suppression chamber bypass leakage is acceptable and continues to meet 10 CFR 50.36(c)(3).

The licensee also proposed deletion of TS 5.5.12, exception c regarding the performance of the next Type A tests as these dates have already occurred and the associated Type A tests have been successfully performed. The NRC staff finds that the deletion of exception c is acceptable because the deleted information is no longer required under 10 CFR 50.36(c)(5).

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Pennsylvania official was notified on August 13, 2025 [21], of the proposed issuance of the amendments. The Commonwealth official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 [3], 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), the NRC staff does not need to prepare an environmental impact statement or environmental assessment 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.

7.0 REFERENCES

[1] Casulli, E., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Proposed Amendment to Licenses NPF-14 and NPF-22: Request to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program, for Permanent Extension of Type A and Type C Leak Rate Testing Frequencies and Permanently Extend the Drywell Bypass Leakage Test Frequency, PLA-8135, November 1, 2024, ADAMS Accession No. ML24306A122.

[2] Casulli, E., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Response to Request for Additional Information To Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program, for Permanent Extension of Type A and Type C Leak Rate Testing Frequencies and, Permanently Extend the Drywell Bypass Leakage Test Frequency, PLA-8182, June 26, 2025, ADAMS Accession No. ML25177C510.

[3] U.S. Nuclear Regulatory Commission, Monthly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, Office of the Federal Register, Volume 90, Page 7193 (90 FR 7193),

January 21, 2025, ADAMS Accession No. ML25002A202.

[4] U.S. Nuclear Regulatory Commission, Regulatory Guide 1.163, Revision 1, PerformanceBased Containment Leak-Test Program, June 2023, ADAMS Accession No. ML23073A154.

[5] Nuclear Energy Institute, NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012, ADAMS Accession Nos. ML12221A202 (report) and ML122210254 (package).

[6] American National Standards Institute/American Nuclear Society, ANSI/ANS 56.8-2020, Containment System Leakage Testing Requirements, December 11, 2020.

[7] U.S. Nuclear Regulatory Commission, Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, September 1995, ADAMS Accession No. ML003740058.

[8] Colburn, T., U.S. Nuclear Regulatory Commission, letter to Byram, R., PPL Susquehanna, LLC, Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Re: One-Time Deferral of Containment Integrated Leak Rate Test and Drywell-to-Suppression Chamber Bypass Leakage Test (TAC Nos. MB2894 and MB2895), March 8, 2002, ADAMS Accession No. ML020280102.

[9] Bahadur, S., U.S. Nuclear Regulatory Commission, letter to Bradley, B., Nuclear Energy Institute, Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), June 8, 2012, ADAMS Accession No. ML121030286.

[10] Nuclear Energy Institute, NEI 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, June 9, 2011, ADAMS Accession Nos. ML112920567 (report) and ML112920570 (package).

[11] Electric Power Research Institute Technical Report 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, October 31, 2008.

[12] Maxin, M., U.S. Nuclear Regulatory Commission, letter to Butler, J., Nuclear Energy Institute, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J and Electric Power Research Institute (EPRI)

Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (TAC No. MC9663), June 25, 2008, ADAMS Accession No. ML081140105.

[13] U.S. Nuclear Regulatory Commission, Regulatory Guide 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, ADAMS Accession No. ML17317A256.

[14] U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, May 2011, ADAMS Accession No. ML100910008.

[15] Poslusny, C., U.S. Nuclear Regulatory Commission, letter to Byram, R., Pennsylvania Power and Light Company, Susquehanna Steam Electric Station, Units 1 And 2 (TAC Nos. M94922 and M92923), September 6, 1996, ADAMS Accession No. ML010110174.

[16] American Society for Testing and Materials, ASTM D5163-08, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants, 2008.

[17] Guzman, R., U.S. Nuclear Regulatory Commission, letter to McKinney, B., PPL Susquehanna, LLC, Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Re: Technical Specification Task Force 343 and 479 (TAC Nos. MD2994 and MD2995), March 19, 2007, ADAMS Accession No. ML070710224.

[18] Casulli, E., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Submittal of Updated Final Safety Analysis Report Revision 71 and Fire Protection Review Report Revision 25 PLA-8081, October 12, 2023, ADAMS Accession No. ML23291A105.

[19] Klett, A., U.S. Nuclear Regulatory Commission, letter to Cimorelli, K., Susquehanna Nuclear, LLC, Susquehanna Steam Electric Station, Units 1 and 2Issuance of Amendment Nos. 282 and 265 Regarding Risk-Informed Completion Times in Technical Specifications (EPID L-2021-LLA-0062), August 30, 2022, ADAMS Accession No. ML22200A062.

[20] Electric Power Research Institute, EPRI Report No. 104285, A Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Palo Alto, California, August 1994.

[21] Edwards, T., U.S. Nuclear Regulatory Commission, email to Sheilds, M., Pennsylvania Department of Environmental Protection, NRC Consultation with Commonwealth of Pennsylvania re. Proposed NSHC for Susquehanna Units 1 & 2 License Amendment Requests, August 14, 2025, ADAMS Accession No. ML25252A210.

[22] U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020, ADAMS Accession No. ML20238B871.

8.0 ABBREVIATIONS AMP aging management program ILRT integrated leak rate test(ing)

ANS American Nuclear Society ISI inservice inspection ANSI American National Standards Institute LAR license amendment request ASME American Nuclear Society LERF large early release frequency ASTM American Society for Testing and Materials LLRT local leak rate test(ing)

BPV boiler and pressure vessel NEI Nuclear Energy Institute CCFP conditional containment failure probability NRC Nuclear Regulatory Commission CDF core damage frequency OM operation and maintenance of nuclear power plants CFR Code of Federal Regulations PRA probabilistic risk assessment CISI containment inservice inspection RG regulatory guide CIV containment isolation valve SR surveillance requirement(s)

DWBT drywell-to-suppression chamber bypass leak rate test TS technical specification(s)

EPRI Electric Power Research Institute UFSAR Updated Final Safety Analysis Report Principal Contributors:

Hanry Wagage, NRR Shaohua Lai, NRR David McClain, NRR Thomas Scarbrough, NRR Joshua Wilson, NRR David Coy, NRR Date: September 23, 2025

ML25253A408

• via eConcurrence NRR-058 OFFICE NRR/DORL/LPL1/PM* NRR/DORL/LPL1/PM* NRR/DORL/LPL1/LA* NRR/DEX/ESEB/BC*

NAME TEdwards AKlett KEntz ITseng DATE 9/11/2025 9/11/2025 9/11/2025 9/12/2025 OFFICE NRR/DEX/EMIB/BC NRR/DRA/APLB/BC NRR/DSS/SCPB/BC NRR/DSS/STSB/BC*

NAME SBailey EDavidson MValentin (BLee for) SMehta DATE 9/22/2025 9/19/2025 8/29/2025 9/19/2025 OFFICE NRR/DORL/LPL1/BC* NRR/DORL/LPL1/PM*

NAME HGonzález AKlett DATE 9/23/2025 9/23/2025