Text

License Amendment and Alternative Request Related to One-Time Exception of Primary Containment Isolation Valve Testing
	ML25282A072
Person / Time
Site:	Limerick
Issue date:	10/09/2025
From:	Para W; Constellation Energy Generation
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LG-25-132
	Download: ML25282A072 (1)
	v • d • e

200 Energy Way Kennett Square, PA 19348 www.constellation.com LG-25-132 10 CFR 50.90 10 CFR 50.55a October 9, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Unit 1 Renewed Facility Operating License No. NPF-39 NRC Docket No. 50-352

Subject:

License Amendment and Alternative Request Related to One-Time Exception of Primary Containment Isolation Valve Testing

Reference:

1.

Public Preapplication Meeting with Constellation Energy Generation, LLC and the NRC on September 29, 2025, to discuss Planned Licensing Request for extending Local Leak Rate and Inservice Tests for Limerick Generating Station, Unit 1 (ADAMS Accession No. ML25262A068)

This letter submits a License Amendment Request (LAR) and an Alternative Request, both of which are related to leakage testing of Primary Containment Isolation Valves (PCIVs).

In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Constellation Energy Generation, LLC (CEG) requests an amendment to Facility Operating License No. NPF-39 for Limerick Generating Station (LGS), Unit 1. The proposed change would modify TS 6.8.4.g Primary Containment Leakage Rate Testing Program to allow a one-time exception to the 10 CFR 50, Appendix J, Type C Local Leakage Rate Tests (LLRTs) required for the four (4)

PCIVs listed in Attachment 1. The proposed change extends the Type C LLRT due date for these valves to no later than the completion of the LGS, Unit 1 Spring 2028 Refueling Outage.

Additionally, in accordance with 10 CFR 50.55a(z)(2), CEG proposes an alternative to the requirement of American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Section ISTC-3630 Leakage Rate for Other Than Containment Isolation Valves for the 3 Pressure Isolation Valves (PIVs) listed in Attachment 3, which are also included in the 4 PCIVs. The Alternative Request is a one-time exception to Relief Request (RR) GVRR-8 to defer resetting the test interval for the 3 valves to allow the next ASME Inservice Testing (IST) program leakage tests to be completed no later than completion of the LGS, Unit 1 Spring 2028 refueling outage.

A preapplication meeting with the NRC was held on September 29, 2025, to discuss this submittal (Reference 1).

This request is divided as follows:

provides a description and evaluation of the proposed LAR.

U.S. Nuclear Regulatory Commission License Amendment and Alternative Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket Nos. 50-352 Page 2

provides a markup of the affected TS page.

provides 10 CFR 50.55a Alternative Request GVRR-12 The proposed changes have been reviewed by the LGS Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

CEG requests approval of the proposed change by March 31, 2026. Once approved, the amendment will be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), CEG is notifying the State of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no new regulatory commitments contained in this submittal.

Should you have any questions concerning this submittal, please contact Lane Oberembt at 267-533-5301.

I declare under the penalty of perjury that the foregoing is true and correct. Executed on the 9th day of October 2025.

Respectfully, Wendi Para Sr. Manager - Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Changes

2. Markup of Proposed Technical Specification Page

3. 10 CFR 50.55a Alternative Request GVRR-12 cc:

USNRC Region I, Regional Administrator w/ Attachments USNRC Senior Resident Inspector, LGS USNRC Project Manager, LGS Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection Respectfully,

ATTACHMENT 1 Limerick Generating Station, Unit 1 Docket No. 50-352 Evaluation of Proposed Changes

Subject:

License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Current and Proposed TS Wording

3.0 TECHNICAL EVALUATION

3.1 System Descriptions 3.2 Chronology of 10 CFR 50, Appendix J Testing Requirements 3.3 Local Leakage Rate Testing Requirements 3.4 Type B and Type C Testing 3.5 Integrated Leakage Rate Testing History 3.6 Defense-in-Depth Assessment 3.7 Justification for Valve Testing Exceptions

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

7.0 FIGURES

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 1 of 32 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), proposes changes to the Technical Specification (TS), Appendix A of Renewed Facility Operating License No. NPF-39 for Limerick Generating Station (LGS), Unit 1.

The proposed change would modify TS 6.8.4.g Primary Containment Leakage Rate Testing Program, to allow a one-time exception of specified 10 CFR 50, Appendix J (Appendix J) Type C Local Leakage Rate Tests (LLRTs). Specifically, TS 6.8.4.g, requires that this program be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing PerformanceBased Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012 (Reference 1), and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008 (Reference 2). For Type C LLRTs, NEI 94-01, Rev. 3-A, Section 10.2.3.4 requires the testing frequency for Primary Containment Isolation Valves (PCIVs) whose test results were not acceptable during the last test, to be reset to the initial test interval of at least once per 30 months. NEI 94-01 has no provisions that allow extensions past the 30-month baseline frequency until sufficient historical as-found test data is available to warrant extending the test interval. The proposed change extends the Type C LLRT due date for four PCIVs listed in Table 3.4-2 to allow these tests to be performed during the LGS, Unit 1 Spring 2028 refueling outage.

The proposed change also includes an administrative change to the format of the LGS, Unit 1 TS 6.8.4.g, Primary Containment Leakage Rate Testing Program.

LGS will be installing the Digital Modernization Project (DMP) in the Spring 2026 refueling outage. The DMP will replace the existing analog control logic hardware of the Reactor Protection System (RPS) instrumentation, Nuclear Steam Supply Shutoff System (NSSSS) instrumentation, the Emergency Core Cooling System (ECCS) instrumentation, the Reactor Core Isolation Cooling (RCIC) System instrumentation, and the End-of-Cycle Recirculation Pump Trip (EOC-RPT) instrumentation with a new single digital control system.

Due to the complex nature of the DMP, the current Unit 1 Spring 2026 refueling outage plan intentionally excludes work on the B train of ECCS. The B train of equipment will have controls set up at alternate locations in the plant, allowing the operators to operate this equipment outside of the Main Control Room (MCR). This is being done to maintain Shutdown Cooling and alternate Reactor Coolant System (RCS) makeup injection sources. The four PCIVs that CEG is requesting deferral of are all located on the B train of equipment.

Performance of the LLRTs on the requested valves introduces significant additional measures that impact nuclear and radiological safety risks.

As currently planned, performing B ECCS LLRTs would also result in a YELLOW defense-in-depth condition on the decay heat removal key safety function due to unavailability of B and D Residual Heat Removal (RHR) trains. A and C RHR trains will be the only available trains for decay heat removal, resulting in an N+1 condition. To achieve GREEN defense in depth of N+2, it would require either B or D RHR train to remain available.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 2 of 25 The LGS DMP is structured around key technical and regulatory advances that have come to fruition in recent years. This structure demonstrates that large scale modernization is a viable economic and technical alternative. Given the scale of the LGS DMP, the public-facing research on which it is founded, and the demonstration of the multiple industry and regulatory initiatives, this project has significant benefits for the industry and, by extension, the public in the form of demonstrating that modernization can be achieved efficiently and will preserve reliable and carbon-free generation. The successful installation strategy being employed with the support of this proposed License Amendment Request, if approved, will facilitate a safe and error-free installation in a timely manner for such a large and significant undertaking.

2.0 DETAILED DESCRIPTION 2.1 Current and Proposed TS Wording LGS, Unit 1 TS 6.8.4.g, Primary Containment Leakage Rate Testing Program, currently states, in part:

Unit 1:

A program shall be established to implement 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 NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44.0 psig.

The proposed change to LGS, Unit 1 TS 6.8.4.g will add an exception to allow for the performance of the next Appendix J, Type C test no later than April 30, 2028 for valves HV-052-108, HV-051-1F015B, HV-051-1F017B, and HV-051-1F027B. In addition, the proposed change adds a second exception to allow the Type C test to be extended indefinitely if the test interval ends while primary containment integrity is not required (i.e., TS 3.6.1, "Primary Containment,"

does not require the primary containment to be operable in Modes 4 and 5). In this case, the second exception requires that the Type C test be performed prior to entering Mode 3. The exceptions are denoted in the proposed change as x and xx in brackets [ ]. This is to indicate that exception numbers will be the next applicable number (i.e., x will be 1 or 3, xx will be 2 or 4) depending on approved amendments at the time of issuance.

An administrative format change would move the last sentence of the LGS, Unit 1 TS to a new paragraph to make the format the same format as the LGS, Unit 2 TS.

The proposed LGS, Unit 1 TS 6.8.4.g will be as follows:

Unit 1:

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 3 of 25 A program shall be established to implement 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 NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008 as modified by the following exceptions: [x] the next Type C test for valves HV-052-108, HV-051-1F015B, HV-051-1F017B, and HV-051-1F027B will be performed no later than April 30, 2028, and [xx] if the Type C test has not been performed by April 30, 2028, and the unit is in Mode 4 or 5, the Type C test shall be performed prior to entering Mode 3.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44.0 psig.

A TS page markup of the proposed change is provided in Attachment 2.

3.0 TECHNICAL EVALUATION

3.1 System Descriptions 3.1.1 Containment Isolation System The containment isolation system is designed to prevent or limit the release of radioactive materials that may result from postulated accidents. This is accomplished by providing isolation barriers in all fluid lines that penetrate primary containment. It is designed with provisions for periodic operability and leak rate testing.

3.1.2 Core Spray System Each of the two redundant Core Spray (CS) system loops consists of: two 50% capacity centrifugal pumps powered from Class 1E buses; a spray sparger in the reactor vessel above the core (a separate sparger for each CS loop); piping and valves to convey water from the suppression pool to the sparger; and associated controls and instrumentation. A connection to the High Pressure Coolant Injection (HPCI) system is provided to allow HPCI injection through the CS Loop B vessel connection.

When low water level in the reactor vessel and/or high pressure in the drywell is sensed, and if reactor vessel pressure is low enough, the CS system automatically starts and sprays water into the top of the fuel assemblies to cool the core. The CS injection piping enters the vessel, divides, and enters the core shroud at two points near the top of the shroud. A semicircular sparger is attached to each outlet. Nozzles are spaced around the sparger to spray the water radially over the core and into the fuel assemblies.

The CS system is designed to provide cooling to the reactor core only when the reactor vessel pressure is low, as is the case for large loss of coolant accident (LOCA) break sizes. However, when CS operates in conjunction with the Automatic Depressurization System (ADS), the

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 4 of 25 effective core cooling capability of CS is extended to all break sizes because the ADS rapidly reduces the reactor vessel pressure to the CS operating range.

3.1.2.1 Penetration X-16B: B Core Spray Discharge (Figure 7.1)

The CS loop B line penetrates the drywell to inject directly into the Reactor Pressure Vessel (RPV). Isolation is provided by two valves in the CS line, a pneumatic testable check valve inside the containment, and a spring assisted check valve outside the containment, with positions of both indicated in the main control room. The core spray loop B line is also provided with a normally closed pneumatic operated globe valve which bypasses the inboard isolation valve for equalization during testing. Although not a containment isolation valve (CIV), a motor operated valve (MOV) is provided outboard of the spring assisted check valve to provide isolation of this flow path for long-term leakage control. One branch of the HPCI line connects to the CS loop B upstream of the outboard spring assisted check valve.

As outlined in Section 6 of the Updated Finals Safety Analysis Report (UFSAR), the inboard boundary is a check valve, HV-052-108, and the outboard boundary is a credited closed system outside of containment. The penetration is 12 inch diameter line. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a High Energy Line Break (HELB), is missile-protected, and is capable of being leak tested.

3.1.3 Residual Heat Removal Shutdown Cooling Two redundant heat exchanger loops are provided to remove residual heat, each with two alignable RHR pumps powered from separate emergency buses. Either heat exchanger loop is capable of cooling down the reactor within a reasonable length of time. During cold shutdown and refueling operation conditions, this results in the availability of four shutdown cooling subsystems. The shutdown cooling suction line from the recirculation suction piping is common to all four shutdown cooling subsystems, and each heat exchanger and its associated discharge piping is common to its two alignable RHR pumps (A and C RHR pumps for the A heat exchanger, B and D RHR pumps for the B heat exchanger).

The functional design basis of the shutdown cooling mode is to have the capability to remove decay and sensible heat from the reactor primary system so that the reactor outlet temperature is reduced to 125ºF, approximately 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the control rods have been inserted, and to permit refueling when the RHR Service Water temperature is 85ºF, assuming that the core is "mature" and the RHR heat exchanger tubes are completely fouled.

3.1.3.1 Penetration X-13B: B RHR Shutdown Cooling Return (Figure 7.1)

Each RHR shutdown cooling return line penetrates primary containment and discharges into a recirculation pump discharge line that injects directly into the RPV. Isolation is provided by an automatically actuated motor-operated globe valve outside containment, in this case HV-051-1F015B, and a pneumatic testable check valve and a spring-assist check valve inside containment. A normally closed pneumatic operated globe valve is provided which bypasses the inboard isolation testable check valve for equalization during testing. To increase the reliability

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 5 of 25 of RHR shutdown cooling mode during refueling outages, the automatic isolation function of the RHR shutdown cooling mode return motor-operated valves is typically bypassed provided that automatic isolation is not required by the TS or Technical Requirements Manual and the reactor cavity is flooded up. Manual isolation capability is retained.

This penetration provides a flow path for the RHR system operating in shutdown cooling mode to return reactor coolant from the RHR heat exchanger to the B reactor recirculation loop.

The isolation provisions for this line consist of one isolation valve outside containment and a closed system outside containment. A single active failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested. The penetration is 12 inch diameter line.

3.1.4 Residual Heat Removal Low Pressure Coolant Injection The Low Pressure Coolant Injection (LPCI) subsystem is an operating mode of the RHR system. The LPCI subsystem is automatically actuated by low water level in the reactor and/or high pressure in the drywell coincident with low reactor pressure. It uses four motor-driven RHR pumps to draw suction from the suppression pool and inject cooling water flow into the reactor core via separate vessel nozzles and core shroud penetrations.

The LPCI subsystem, like the CS system, is designed to provide cooling to the reactor core only when the reactor vessel pressure is low, as is the case for large LOCA break sizes. However, when LPCI operates in conjunction with the ADS, the effective core cooling capability of LPCI is extended to all break sizes because the ADS rapidly reduces the reactor vessel pressure to the LPCI operating range.

3.1.4.1 Penetration X-45B: B RHR LPCI (Figure 7.1)

The LPCI lines are provided with remote manually controlled gate valves, in this case HV-051-1F017B, outside and pneumatic testable check valves inside containment. Both types of valves are normally closed with the gate valves receiving an automatic signal to open at the appropriate time. The check valves are located as close as practicable to the RPV. The normally closed check valves limit containment pressurization if there is a pipe rupture between the check valve and containment wall. The LPCI lines are also provided with a normally closed pneumatic operated globe valve which bypasses the inboard isolation valve for testing purposes.

This penetration provides a flow path for the B LPCI. The isolation provisions for this line consist of one isolation valve outside containment and a closed system outside containment. A single active failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested. The penetration is 12 inch diameter line.

3.1.5 Residual Heat Removal Suppression Pool Spray

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 6 of 25 The RHR system is comprised of four independent loops. Each loop takes suction from the suppression pool and is capable of discharging water to the reactor for low pressure coolant injection vessel via a separate vessel nozzle or back to the suppression pool via a full flow test line.

3.1.5.1 Penetration X-205B: B RHR Suppression Pool Spray (Figure 7.2)

The suppression pool spray lines are provided with normally closed isolation valves outside containment, located directly on the containment. The valves automatically close upon receipt of an isolation signal. The external pipe, designed to Quality Group B and seismic Category I requirements, provides the second isolation barrier. Because of the desired use of this system after a LOCA, the system reliability is greater with only one isolation valve in the line.

This penetration provides a flow path for the suppression pool spray mode of RHR. The isolation provisions for this line consist of one isolation valve outside containment, in this case HV-051-1F027B, and a closed system outside containment. A single active failure can be accommodated. The closed system is missile-protected, seismic Category I, quality group B and designed to the temperature and pressure conditions that the system will encounter post-LOCA.

All isolation barriers are located outside containment.

3.2 Chronology of Testing Requirements of 10 CFR 50, Appendix J The testing requirements of Appendix J provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. Appendix J also ensures that periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment and the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests: (1) Type A tests, intended to measure the primary containment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations, and (3) Type C tests, intended to measure containment isolation valve leakage rates. Type B and C tests identify most potential containment leakage paths.

Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Type B and C testing.

In 1995, Appendix J was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach.

Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 7 of 25 Regulatory Guide (RG) 1.163, Section C, Regulatory Position," (Reference 3) states that NEI 94-01, Revision 0, dated July 26, 1995, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J," (Reference 4) 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 conditions specified in RG 1.163, Section C. NEI 94-01 provides the criteria for LLRT frequency extensions based on containment component performance. For Type C LLRTs, NEI 94-01, provided guidance for an extended interval of greater than 60 months. However, RG 1.163 did not endorse this recommendation and limited the maximum Type C test interval to 60 months.

In 2008, NEI 94-01, Revision 2-A (Reference 2), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation (SE) on NEI 94-01 (Reference 5). The major change in NEI 94-01, Revision 2-A, included provisions for extending Type A Integrated Leakage Rate Tests (ILRT) intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

In 2012, NEI 94-01, Revision 3-A (Reference 1), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to Appendix J and includes provisions for extending Type C LLRT intervals to up to a maximum of 75 months. NEI 94-01 has been endorsed by RG 1.163 and NRC Safety Evaluations (SEs) dated June 25, 2008 (Reference 5) and June 8, 2012 (Reference 6) as an acceptable methodology for complying with the provisions of Option B in Appendix J. The regulatory positions stated in RG 1.163 as modified by References 5 and 6 are incorporated in this document. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

3.3 Local Leakage Rate Testing Requirements On March 11, 2020, the NRC issued Amendment 241 for LGS, Unit 1 (Reference 7). This amendment approved an extension of the containment isolation valve leakage rate testing frequency from 60 months to 75 months for Type C leakage rate testing of selected components in accordance with NEI 94-01, Revision 3-A. This revision also restated the Type C testing requirements as outlined below.

3.3.1 NEI 94-01, Revision 3-A Test Requirements Section 10.2.3.1 Initial Test Interval Periodic Type C tests shall be performed at a frequency of at least once per 30 months, until adequate performance has been established consistent with Section 10.2.3.2.

Section 10.2.3.2 Extended Test Interval Test intervals for Type C valves may be increased based upon completion of two consecutive periodic as-found Type C tests where the result of each test is within a licensees allowable administrative limits. Intervals for Type C testing may be increased

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 8 of 25 to a specific value in a range of frequencies from 30 months up to a maximum of 75 months.

Section 10.2.3.3 Repairs or Adjustments An as-left Type C test shall be performed following maintenance, repair, modification or adjustment activity unless an alternate testing method or analysis is used to provide reasonable assurance that such work does not affect a valves leak tightness and a valve will still perform its intended function. If as-found and as-left Type C test results are both less than a valves allowable administrative limit, a change of the test frequency is not required. If as-found or as-left test results are greater than the allowable administrative limit, then provisions of Section 10.2.3.4 apply. Testing shall continue at this frequency until an adequate performance history is established in accordance with Section 10.2.3.2.

Section 10.2.3.4 Corrective Action If Type C test results are not acceptable, then the testing frequency should be set at the initial test interval per Section 10.2.3.1. In addition, a cause determination should be performed and corrective actions identified that focus on those activities that can eliminate the identified cause of a failure with appropriate steps to eliminate recurrence.

Once the cause determination and corrective actions have been completed, acceptable performance may be reestablished and the testing frequency returned to the extended interval in accordance with Section 10.2.3.2.

3.4 Type B and Type C Testing A significant number of Type B and Type C containment penetrations are local leak-rate tested during each refueling outage in accordance with the performance-based testing requirements of Appendix J, Option B, as modified by approved exemptions. As stated in TS 6.8.4.g for Unit 1, this testing is in accordance with the guidelines contained in NEI 94-01, Revision 3-A, dated July 2012, and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008.

Type C testing is performed on Containment Isolation Valves in those piping systems which could become a potential leakage pathway from inside the Primary Containment to the outside environs following various accident scenarios up to and including the Design Basis Accident (DBA). Type B testing is performed on Primary Containment penetrations which do not normally see flow, such as personnel and equipment air locks, equipment hatches, and electrical penetrations. All Primary Containment penetrations are tested at a pressure greater than or equal to the peak calculated internal pressure for the design basis accident, Pa (44.0 psig), unless otherwise specifically stated in the Primary Containment Leakage Rate Testing Program.

The Type B and C acceptance criteria are based on running totals of the cumulative leakage rates for all Type B and C penetrations. Acceptance criteria for Type B and C penetrations is less than or equal to 0.60 La, (94,964 sccm) min path, when the Primary Containment is required to be operable and 0.60 La max path, prior to entering a mode requiring the Primary Containment to be operable following testing in accordance with this Program. The testing

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 9 of 25 frequency of Type B and C components is based on performance and may be extended from the base frequency of 30 months to a frequency of 120 months for Type B tests and 75 months for Type C tests. Certain components, such as MSIVs, Feedwater check valves, and Primary Containment Vent and Purge Valves are excluded from performance-based consideration and must be tested on the base 30-month frequency. LGS TS 3.6.1.2.c limits MSIV total leakage not to exceed 200 SCFH (94,389.5 sccm) for all four main steam lines, when tested at 22.0 psig.

Table 3.4-1: Unit 1 Types B & C Performance History REFUELING OUTAGE MSIV AS-FOUND MIN PATH MSIV AS-LEFT MAX PATH TYPES B&C AS-FOUND MIN-PATH TYPE B&C AS-LEFT MAX-PATH Spring 2024 26.12 SCFH 121.58 SCFH 91,564.1 sccm 51,153.8 sccm Spring 2022 87.77 SCFH 102.97 SCFH 47,188.2 sccm 70,384.2 sccm Spring 2020 96.40 SCFH 146.63 SCFH 25,447.0 sccm 57,058.1 sccm The above table demonstrates that the Type B and Type C primary containment leakage rate testing program is well maintained and has a reasonably large margin with regard to the overall Primary Containment Leakage Rate Testing Program.

Installation of the DMP will require the affected components scheduled for Type B and Type C testing to have their test procedures revised to account for the new digital control system being installed versus the existing analog system. There will be no changes to the actual valve or penetration configurations in the plant, or test methodology. LGS does not foresee any instances where the DMP will impact compliance with the requirements of ANSI/ANS 56.8 Section 3.3.3 (Reference 8).

3.4.1 Review of Test History for Valves Requesting Testing Exceptions Currently, the overall MaxPath Leakage for Unit 1 is 51,153.8 sccm, which is equivalent to 0.32La. The penetrations for which deferral is being requested have a combined leakage of 9,670 sccm, which contributes to 19% of the overall MaxPath Leakage, and is equivalent to 0.061La. The margin to 0.6La is 43,809 sccm on a MaxPath basis.

Table 3.4-2: Valves Appendix J, Type C Performance History Refueling Outage Penet. X-16B HV-052-108 Ad. Lim: 3000 Penet. X-13B HV-051-1F015B Ad Lim: 5000 Penet. X-45B HV-051-1F017B Ad Lim: 5000 Penet. X-205B HV-051-1F027B Ad Lim: 5000 Spring 2024 As-Found: Off-scale*

As-Left: 150 sccm 5120 sccm As-Found: 18000 sccm As-Left: 100 sccm As-Found: 10000 sccm As-Left: 4300 sccm Spring 2022 32.5 sccm 3300 sccm 8225 sccm 9750 sccm Spring 2020 Note 1 Note 1 Note 1 7100 sccm Spring 2018 Note 1 Note 1 Note 1 Note 1 Spring 2016 20 sccm 102 sccm 443 sccm 4660 sccm Spring 2014 20 sccm 1162.2 sccm 20 sccm 3720 sccm

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 10 of 25

- Unable to establish test pressure during PIV IST Leakage test, therefore conservatively assigned off-scale as Appendix J leakage.

Note 1 - Penetration was not tested due to being on an extended testing frequency.

Three of the valves for which deferral is being requested, are also subject to American Society of Mechanical Engineers (ASME) Inservice Testing (IST) leakage tests. The test history for these valves is shown below in table 3.4-3.

Table 3.4-3: Valves ASME IST Leakage Performance History Refueling Outage Penet. X-16B HV-052-108 Limit: < 1gpm Penet. X-13B HV-051-1F015B Limit: < 1gpm Penet. X-45B HV-051-1F017B Limit: < 1gpm Spring 2024 As-Found: Off-scale*

As-Left: 0.01 gpm 0.0 gpm As-Found: 0.08 gpm As-Left: 0.01 gpm Spring 2022 0.0 gpm 0.17 gpm 0.0 gpm Spring 2020 Note 1 Note 1 Note 1 Spring 2018 Note 1 Note 1 Note 1 Spring 2016 0.0 gpm 0.1 gpm 0.0 gpm Spring 2014 0.0 gpm 0.0 gpm 0.0 gpm

- Unable to establish test pressure, therefore assigned off-scale as leakage.

Note 1 - Penetration was not tested due to being on an extended testing frequency 3.4.1.1 Penetration X-16B: B Core Spray Discharge (HV-052-108)

During the Spring 2024 refueling outage, the as-found leakage of HV-052-108 was not determined due to the inability to establish test conditions due to high leakage for both tests. In-body maintenance was performed on the valve, and the as-left leakage was measured at 150 sccm, which is 5% of the 3,000 sccm administrative limit. The as-left ASME Pressure Isolation Valve (PIV) IST leakage was 0.01 gpm, which is 1% of the allowable leakage. The Appendix J program requires leakage testing to return to base frequency if the component leakage exceeded the administrative limit and maintenance was performed to return the leakage to below the limit. Therefore, testing in the Spring 2026 refueling outage is required.

3.4.1.2 Penetration X-13B: B RHR Shutdown Cooling Return (HV-051-1F015B)

During the Spring 2024 refueling outage, the as-found leakage of the penetration was found to be 5,120 sccm. A technical evaluation was performed to defer repairs, and no in-body maintenance was performed during the Spring 2024 refueling outage. Therefore, the as-left leakage rate is 5,120 sccm, 102.4% of the 5,000 sccm administrative limit. The ASME PIV IST leakage was 0.0 gpm. The Appendix J program requires a component to return to base frequency if leakage exceeded the administrative limit. Therefore, testing in the Spring 2026 refueling outage is required.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 11 of 25 3.4.1.3 Penetration X-45B: B RHR LPCI (HV-051-1F017B)

During the Spring 2024 refueling outage, the as-found leakage measured at this penetration was 18,000 sccm. In-body maintenance was performed on HV-051-1F017B, and the as-left leakage was measured at 100 sccm, which is 2% of the 5,000 sccm administrative limit. The ASME PIV IST as-found leakage testing was measured at 0.08 gpm, which is 8% of the allowable leakage. After the in-body maintenance was performed, the ASME PIV IST as-left leakage was measured at 0.01 gpm, which is 1% of the allowable leakage. The Appendix J program requires leakage testing to return to base frequency if the component leakage exceeded the administrative limit and maintenance was performed to return the leakage to below the limit. Therefore, testing in the Spring 2026 refueling outage is required.

3.4.1.4 Penetration X-205B: B RHR Suppression Pool Spray (HV-051-1F027B)

During the Spring 2024 refueling outage, the as-found leakage at this penetration was 10,000 sccm. In-body maintenance was performed on HV-051-1F027B, and the as-left leakage was measured at 4,300 sccm, which is 86% of the 5,000 sccm administrative limit. Two linear indications were noted on the Stellite hardfacing of the valve seat. The Appendix J program requires leakage testing to return to base frequency if the component leakage exceeded the administrative limit and maintenance was performed to return the leakage to below the limit.

Therefore, testing in the Spring 2026 refueling outage is required.

3.5 Integrated Leakage Rate Testing History Previous LGS ILRT results have confirmed the containment is acceptable, with considerable margin, with respect to the TS acceptance criterion of 0.5% of primary containment air weight per day at the design basis loss-of-coolant accident pressure.

Table 3.5 LGS Unit 1 Type A Testing History Test Date 95% UCL (wt.%/day)

Note 3 As-Found Leakage (wt.%/day)

Acceptance Criteria(wt.%/day)

As-Left Leakage (wt.%/day)

As-Found (La)

As-Left (75% La) 8/3/1984 0.213 Note 1 Note 1 0.375 0.1642 8/13/1987 0.131 Note 2 0.5 0.375 0.1469 11/23/1990 0.252 Note 4 0.5 0.375 0.287 5/13/1998 0.263 0.3751 0.5 0.375 0.307 3/17/2012 0.139 0.2688 0.5 0.375 0.2318 Note 1: This was a pre-operational test; therefore, no as-found leak rate calculated.

Note 2: The as-found test results failed to meet the acceptance criteria of 0.500wt.%/day.

Note 3: The upper confidence limit (UCL) is a calculated value determined from test data that places a statistical upper bound on the true leakage rate. The UCL is calculated at a 95% confidence level in ANSI/ANS 56.8. From this 95% UCL leakage rate value, both

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 12 of 25 the as-left and the as-found ILRT leakage rates are determined. Corrections are made to the 95% UCL leakage rate for changes in the net free volume due to changes in containment sub-volume water levels and valves not in accident positions (Types B and C penalties) during the test.

Note 4: LGS does not maintain records of Types B and C leak rate summations for refueling outages earlier than 1996. Therefore, leakage savings are not known, and the as-found leak rate cannot be calculated.

Performed at same time as ILRT was the Drywell Bypass Leak Rate Test (DWBT) assessment, results of which have demonstrated acceptable margin to the leakage limit. The history of test results indicates that the typical leakage is about an order of magnitude or more below the acceptance criteria (which is set at an order of magnitude below the design basis limit).

Table 3.5 Unit 1 DWBT Historical Results Year Measured Leakage (ft2)

Acceptance Criteria (ft2) 1984 0.00026 0.005 1987 0.00005133 0.005 1990 0.000278 0.005 1998 0.000075 0.005 2012 0.000151 0.005 3.6 Defense-in-Depth Assessment LGS has testing programs and associated inspections, in addition to the Primary Containment Leakage Rate Testing Program, that are designed to identify any degrading conditions and ensure the containment structure remains capable of meeting its design functions. These programs include:

1. Protective Coating Program - A program to maintain containment coatings was developed to meet the requirements of RG 1.54 Rev. 0. This program is implemented in accordance with Constellation Procedure ER-AA-330-008 "Safety-Related (Service Level I) Protective Coatings". This includes the Drywell and Suppression Pool coatings which protect the integrity of the containment lining.

2. Containment Inservice Inspection Program (CISI) - LGS performs a comprehensive primary containment inspection to the requirements of ASME Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Subsection IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants," and Subsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Plants." These include the Containment Liner, Concrete Containment, and Containment Penetrations, both inside and outside. These inspections ensure the structural integrity of containment.

3.7 Justification for Valve Testing Exceptions

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 13 of 25 Due to the complex nature of the DMP, the current the Spring 2026 refueling outage plan intentionally excludes work on the B train of the ECCS. This was planned due to the alternate controls scheduled to be installed on the B ECCS equipment at the start of the outage in order to maintain adequate sources of Shutdown Cooling and Alternate Injection Sources.

If B ECCS LLRTs are required, it would introduce significant additional measures that carry nuclear and radiological safety risks. If B ECCS LLRTs are required, there are two different options for the outage.

1. The first option is to shut down the plant on the A train of RHR Shutdown Cooling using normal controls from MCR. At approximately 55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months into the outage, the station would then perform a swap from the normal controls on the A shutdown cooling train to the alternate controls for this same train which will need to be installed at the Remote Shutdown Panels. This is being done to maintain outage safety and would require another Shutdown Cooling system swap mid outage. This would require four additional Motor Operated Valve (MOV) test boxes and three more 4kV Breaker test boxes. Three additional system indicator hook-ups in the field would be required, along with new temporary RHR Heat Exchanger controller wiring. This approach also introduces additional unplanned terminations and complicates B ECCS system refill under interim operating conditions due to additional alternate controls to support the fill and vent. This additional work to prepare the A train would also incur added dose to the workers. It also necessitates a complex and infrequently performed evolutionswapping shutdown cooling trains from A ECCS (on alternate controls) to B ECCS (also on alternate controls). This would require multiple teams in multiple locations in different enclosures with complicated communications. While operators are trained for this task, the dual use of alternate controls increases the risk profile unnecessarily.

2. Alternatively, a full core offload could be pursued, requiring the removal of every fuel bundle from the reactor core to the spent fuel pool. This introduces additional reactivity management challenges due to the additional fuel moves, which is not justified given the original outage scope.

In either case, 500mR of accumulated dose is expected to perform the additional work and applicable leakage testing (LLRT and PIV IST).

In summary, performing B ECCS LLRTs during the Spring 2026 refueling outage introduces avoidable risks and operational complexity, and the safest path forward is to adhere to the original plan and exclude B ECCS testing from the Spring 2026 refueling outage.

3.7.1 Penetration X-16B: B Core Spray Discharge (HV-052-108)

HV-052-108 is a 12-inch Anchor Darling power assisted swing check valve. Appendix J, Type C leakage testing is required in the Spring 2026 refueling outage per TS 6.8.4.g, as implemented by NEI 94-01, since the valve failed its as-found Appendix J, Type C leakage test in Spring 2024 refueling outage. This valve also failed its freedom of movement test in the 2024 refueling outage. Based on the as-found in-body inspection of the valve, the failure is suspected to be a

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 14 of 25 combination of the valve potentially not traveling to the seat under test conditions (slow pressure increase) along with normal in-body wear. The valve received maintenance to correct the freedom of movement issue, and the in-body seat and disc were reconditioned with a satisfactory as-left Appendix J, Type C leakage test.

Extension of the LLRT is requested until the Spring 2028 refueling outage due to the valves strong Appendix J leakage history before the 2024 failure (Table 3.4-2), recent maintenance, and satisfactory as-left leakage results. The HV-052-108 valve had performed reliably for over 16 years without requiring corrective maintenance. This valve has not had recent maintenance, with subsequent LLRT history, due to its satisfactory performance for over 14 years. Limerick utilizes robust maintenance procedures and practices performed by qualified technicians to perform valve repairs. From 2008 until 2024, LLRT results were consistently less than 100 sccm, with only one failure in 2024. Based on the refurbishment and reconstitution of the disc and seat and its stable historical performance, the valve is expected to have reliable leakage performance through the Spring 2028 refueling outage.

Additionally, the second penetration isolation barrier is a closed system outside containment. A single active valve failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested.

The HV-052-108 will also be exercise tested in the Spring 2026 refueling outage to demonstrate functionality and freedom of movement.

3.7.2 Penetration X-13B: B RHR Shutdown Cooling Return (HV-051-1F015B)

HV-051-1F015B is a 12-inch Anchor Darling Motor Operated Globe Valve. Appendix J, Type C leakage testing is required in the Spring 2026 refueling outage per TS 6.8.4.g, as implemented by NEI 94-01, since the valve failed its as-found Appendix J, Type C leakage test in Spring 2024 refueling outage.

Extension of the Appendix J, Type C testing is requested until the Spring 2028 refueling outage due to the valves historically satisfactory LLRT test performance (Table 3.4-2). The HV-051-1F015B valve has operated reliably for over 16 years without requiring corrective maintenance.

This valve has had acceptable Appendix J, Type C leakage tests for 16 years with the most recent test result in 2024 of 5,120 sccm. The measured leakage is only 2.4% over the administrative limit of 5,000 sccm. While this is considered a failure per NEI 94-011 for exceeding the administrative limit, the measured Appendix J, Type C leakage is acceptable as compared to the overall 0.6 La limit. From 2008 to 2022, measurements have indicated that Appendix J, Type C leakage rates were consistently less than 3,300 sccm. Given this history of acceptable Appendix J, Type C leakage performance, the valve is expected to have acceptable leakage performance through the 2028 outage.

1 Per NEI 94-01, a failure in this context is exceeding an administrative limit and not the total failure of the penetration. Administrative limits are established at a value low enough to identify and allow early correction of potential total penetration failures.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 15 of 25 Additionally, the second penetration isolation barrier is a closed system outside containment. A single active valve failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested.

HV-051-1F015B is an Active MOV and therefore tested per the ASME OM Appendix III Program. This valve is a Low Risk, High Margin valve that is diagnostically tested. The last diagnostic test was performed in the 2024 refueling outage. There were no anomalies identified during this test that would adversely impact valve functionality, and the results ensure reliable performance and consistent seating force for acceptable leakage performance through 2028.

HV-051-1F015B will be exercised and timed in the Spring 2026 refueling outage to demonstrate functionality and ability to meet the required isolation time.

3.7.3 Penetration X-45B: B RHR LPCI (HV-051-1F017B)

HV-051-1F017B is a 12-inch Anchor Darling Motor Operated Gate Valve. Appendix J, Type C leakage testing is required in the Spring 2026 refueling outage per TS 6.8.4.g, as implemented by NEI 94-01, since the valve failed its as-found Appendix J, Type C leakage test in Spring 2024 refueling outage. This valve exhibited high Appendix J, Type C leakage in 2024 refueling outage and in-body maintenance was performed. Based on the as-found in-body inspection of the valve, the high leakage was attributed to indications on the seat and normal wear. The in-body seat and disc were reconditioned removing the indication and the valve had satisfactory as-left Appendix J, Type C and PIV IST leakage test results. The as-left Appendix J, Type C leakage following maintenance was 100 sccm.

Extension of Appendix J, Type C leakage testing is requested until the Spring 2028 refueling outage due to the valves historically satisfactory performance (Table 3.4-2). In 2012, this valve experienced elevated Appendix J, Type C leakage due to a planned change in MOV thrust resulting in elevated leakage prompting inbody maintenance. The 2012 as-left Appendix J, Type C leakage test result was 20 sccm. The two subsequent Appendix J, Type C leakage results following the 2012 maintenance were less than 450 sccm. This historical performance demonstrated Limericks robust maintenance procedures and practices and provides confidence in the valves leakage integrity following maintenance. Given the stable history of acceptable Appendix J, Type C leakage performance following 2012 maintenance, the valve is expected to have reliable leakage performance following the 2024 inbody maintenance, through the 2028 outage.

Additionally, the second penetration isolation barrier is a closed system outside containment. A single active valve failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested.

HV-051-1F017B is an Active MOV and therefore tested per the ASME OM Appendix III Program. This valve is a Medium Risk, High Margin valve that is diagnostically tested. The last diagnostic test was performed following the inbody repair in 2024. There were no anomalies identified during this test that would adversely impact valve functionality, and the

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 16 of 25 results ensure reliable performance and consistent seating force for acceptable leakage performance through 2028. HV-051-1F017B will be exercised and timed in the Spring 2026 refueling outage to demonstrate functionality and ability to meet the required isolation time.

3.7.4 Penetration X-13B: B RHR Suppression Pool Spray (HV-051-1F027B)

HV-051-1F027B is a 6 Anchor Darling Globe Motor Operated Valve. Appendix J, Type C leakage testing is required in the Spring 2026 refueling outage per TS 6.8.4.g, as implemented by NEI 94-01, since the valve failed its as-found Appendix J, Type C leakage test in Spring 2024 refueling outage. This valve exhibited high Appendix J, Type C leakage in 2024 refueling outage and in-body maintenance was performed. Based on the as-found in-body inspection of the valve, the high leakage was attributed to two indications on the stellite in-body seat. There is no documented previous in-body inspection and therefore it is unknow how long the seat indications existed. The in-body seat and disc were reconditioned; however, the indications could not be entirely removed from the seating surface. The as-left Appendix J, Type C leakage following maintenance was 4,300 sccm which is less than the administrative limit of 5,000 sccm.

A similar indication was observed on the HV-051-2F027A disc in 2011 which is the same make and model valve. This condition was deemed acceptable based on visual inspection and operating experience of Main Steam Isolation Valves (MSIVs) at LGS that had similar stellite indications. The conclusion was that based on engineering judgement and operating experience of MSIVs that are in more severe service, the valve would have acceptable leakage for one cycle. The HV-051-2F027A has had acceptable leakage performance to date, below the administrative limit since the in-body work and identification of the stellite indication in 2011.

Similar performance is expected from HV-051-1F027B.

The historical performance of the HV-051-2F027A sister valve demonstrates Limericks robust maintenance procedures and practices and provides confidence in the HV-051-1F027B valves leakage integrity following maintenance despite the seat indication. Given that the valves leakage performance was reduced to below the administrative limit and the history of acceptable Appendix J, Type C leakage performance of the sister valve following maintenance, the HV-051-1F027B is expected to have reliable leakage performance following the 2024 inbody maintenance, through the 2028 outage.

Additionally, the second penetration isolation barrier is a closed system outside containment. A single active valve failure can be accommodated. The closed system does not communicate with the outside atmosphere, meets Seismic Category I and Safety Class 2 design requirements, designed to temperature and pressure conditions that the system will encounter post-LOCA, is protected from a HELB, is missile-protected, and is capable of being leak tested.

HV-051-1F027B is an Active MOV and therefore tested per the ASME OM Appendix III Program. This valve is a Low Risk, High Margin valve that is diagnostically tested. The last diagnostic test was performed following the inbody repair in 2024. There were no anomalies identified during this test that would adversely impact valve functionality, and the results ensure reliable performance and consistent seating force for acceptable leakage performance through 2028. HV-051-1F027B will be exercised and timed in the Spring 2026 refueling outage to demonstrate functionality and ability to meet the required isolation time.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 17 of 25

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 Regulations 10 CFR 50.36 sets forth the regulatory requirements for the content of the TSs. This regulation requires, in part, that the TS contain Surveillance Requirements (SR). 10 CFR 50.36(c)(3),

states that SRs to be included in the TS are those relating to test, calibration, or inspection which assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the TS LCO will be met.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." 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 which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained.

The one-time extension to the testing frequency of Type C testing will not affect the design, fabrication, or construction of the containment structure or its penetrations. The containment Type C test will continue to be done in accordance with 10 CFR 50 Appendix J using 10 CFR 50 Appendix B quality standards. Although exception to the retest requirements outlined in NEI 94-01, Rev. 3-A is being requested, it is anticipated that overall containment leakage will continue to be acceptable while minimizing unnecessary nuclear safety risks. Therefore, there will be no instances where the applicable regulatory criteria are not met.

10 CFR 50.55a(f) Preservice and inservice testing requirements requires, in part, that Inservice Testing (IST) of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda incorporated by reference in the regulations. Valves HV-052-108, HV-051-1F015B, and HV-051-1F017B will require relief requests to be submitted in accordance with 10 CFR 50.55a(z).

4.1.2 General Design Criteria The LGS UFSAR Section 3.1 discusses the extent to which the design criteria for the plant structures, systems, and components important to safety meet the General Design Criteria (GDC) for Nuclear Power Plants specified in 10CFR50, Appendix A.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 18 of 25 GDC 50 - Containment Design Basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system, shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA.

This margin shall reflect consideration of 1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degraded emergency core cooling functioning, 2) the limited experience and experimental data available for defining accident phenomena and containment responses, and 3) the conservatism of the calculational model and input parameters.

GDC 52 - Capability for Containment Leakage Rate Testing The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

GDC 55 - Reactor Coolant Pressure Boundary (RCPB) Penetrating Containment Each line that is part of the RCPB and that penetrates the primary reactor containment shall be provided with containment isolation valves as follows unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

a. One locked closed isolation valve inside and one locked closed isolation valve outside containment, or

b. One automatic isolation valve inside and one locked closed isolation valve outside containment, or

c. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or

d. One automatic isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements (such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 19 of 25 containment) shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

GDC 56 - Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates the primary reactor containment shall be provided with containment isolation valves, as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

a. One locked closed isolation valve inside and one locked closed isolation valve outside containment, or

b. One automatic isolation valve inside and one locked closed isolation valve outside containment, or

c. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or

d. One automatic isolation valve inside and one automatic isolation valve outside containment.

A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to the containment as practical and, upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

The LGS UFSAR Section 6.2.4.3.1.2 outlines variances to GDC 55 for HV-052-108 Loop B CS Discharge, HV-051-1F015B Loop B RHR Shutdown Cooling Return, and HV-051-1F017B Loop B RHR LPCI. LGS UFSAR Section 6.2.4.3.1.3 outlines variance to GDC 56 for HV-051-1F027B Loop B RHR Suppression Pool Spray. These exceptions have been evaluated as acceptable as discussed in the UFSAR, and no additional actions are required.

CEG has determined that the proposed change requires the TS change as submitted with the accompanying relief request. These will be processed in accordance with applicable regulatory and CEG requirements.

4.2 Precedent The proposed amendment incorporates into the LGS, Unit 1 TS a change that is similar to the following license amendment previously approved by the NRC to extend the Type C test frequency:

Watts Bar Nuclear Plant, Unit 2, Issuance of Amendment Regarding One-Time Extension of 10 CFR Part 50, Appendix J, Type C Local Leakage Rate Tests, dated May 18, 2017 (ML17123A228)

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 20 of 25 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), proposes changes to the Technical Specification (TS), Appendix A of Renewed Facility Operating License No. NPF-39 for Limerick Generating Station (LGS), Unit 1.

The proposed change would modify TS 6.8.4.g Primary Containment Leakage Rate Testing Program, to allow a one-time exception of specified 10 CFR 50, Appendix J (Appendix J) Type C Local Leakage Rate Tests (LLRTs). Specifically, TS 6.8.4.g, requires that this program be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing PerformanceBased Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012 (Reference 1), and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008 (Reference 2). For Type C LLRTs, NEI 94-01, Rev. 3-A, Section 10.2.3.4 requires the testing frequency for Primary Containment Isolation Valves (PCIVs) whose test results were not acceptable during the last test, to be reset to the initial test interval of at least once per 30 months. NEI 94-01 has no provisions that allow extensions past the 30-month baseline frequency until sufficient historical as-found test data is available to warrant extending the test interval. The proposed change extends the Type C LLRT due date for four PCIVs to allow these tests to be performed during the LGS, Unit 1 Spring 2028 refueling outage.

CEG has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change involves changes to the LGS, Unit 1 Primary Containment Leakage Rate Testing Programs, specifically, the LLRT program. The proposed change does not involve a physical change to the plants or a change in the manner in which the plants are operated or controlled. The primary containment function is to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself, and the testing requirements to periodically demonstrate the integrity of containment, exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators.

The proposed change modifies TS 6.8.4.g, to allow for a one-time exception to the retest requirements for specific Type C components. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 21 of 25

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change modifies TS 6.8.4.g, to allow for a one-time extension to the containment Type C test interval for the specified PCIVs. Containment, and the testing requirements to periodically demonstrate the integrity of containment, exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators. The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC) or a change in the way any SSC is operated. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change modifies TS 6.8.4.g, to allow for a one-time exception to the Type C test requirements outlined in NEI 94-01, Rev. 3-A for specified PCIVs. The proposed change does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. Furthermore, any increase in leakage because of the extension is expected to be within TS limits and will not compromise containment integrity. Extending these LLRTs does not involve a change to any limit on accident consequences specified in the license or regulations, a change in a methodology used to evaluate consequences of an accident, a change in any operating procedure or process, a change in how accidents are mitigated or a significant increase in the consequences of an accident. Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that containment would not degrade in a manner that is not detectable by a Primary Containment Leakage Rate Testing Program.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, CEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 22 of 25 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

CEG has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 23 of 25

6.0 REFERENCES

1. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012.

2. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 2008.

3. NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, Revision 0, September 1995.

4. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 21, 1995.

5. NRC Safety Evaluation Report, 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.

6. NRC Safety Evaluation Report, Safety Evaluation by the Office of Nuclear Reactor Regulation Nuclear Energy Institute Topical Report 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J Nuclear Energy Institute Project No. 689, June 8, 2012.

7. NRC Safety Evaluation Report, Limerick Generating Station, Units 1 and 2 Issuance of Amendment nos. 241 and 204 to Revise Technical Specification 6.8.4.g, Primary Containment Leakage Rate Testing Program, to Extend Containment Integrated Leak Rate Test Frequency (EPID L-2019-LLA-0073), dated March 11, 2020.

8. ANSI/ANS 56.8 - 1994, American National Standard for Containment System Leakage Testing Requirements

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 24 of 25 7.0 FIGURES

U.S. Nuclear Regulatory Commission License Amendment Request Related to One-Time Exception of Primary Containment Isolation Valve Testing Docket No. 50-352 : Evaluation of Proposed Changes Page 25 of 25

ATTACHMENT 2 Markup of Proposed Technical Speci"cations Page Limerick Generating Station, Unit 1 Renewed Facility Operating License No. NPF-39 NRC Docket No. 50-352 UNIT 1 REVISED TECHNICAL SPECIFICATIONS PAGE 6-14c

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) g.

Primary Containment Leakage Rate Testing Program A program shall be established to implement 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 NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the Limitations and Conditions specified in NEI 94-01, Revision 2-A, dated October 2008. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44.0 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.5%

of primary containment air weight per day.

Leakage rate acceptance criteria are:

a.

Primary Containment leakage rate acceptance criterion is less than or equal to 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.60 La for the Type B and Type C tests and less than or equal to 0.75 La for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall airlock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa.

2)

Seal leakage rate is less than or equal to 5 scf per hour when the gap between the door seals is pressurized to 10 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the tests described in the Primary Containment Leakage Rate Testing Program.

h.

Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

A change in the TS incorporated in the license; or A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

LIMERICK - UNIT 1 6-14c Amendment No. 118, 162, 190, 241 Insert A Start new paragraph

Insert A as modi"ed by the following exceptions: [x] the next Type C test for valves HV-052-108, HV-051-1F015B, HV-051-1F017B, and HV-051-1F027B will be performed no later than April 30, 2028, and [xx] if the Type C test has not been performed by April 30, 2028, and the unit is in Mode 4 or 5, the Type C test shall be performed prior to entering Mode 3.

ATTACHMENT 3 Limerick Generating Station, Unit 1 Docket No. 50-352 10 CFR 50.55a Alternative Request GVRR-12

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 1 of 6

1. ASME Code Component(s) Affected Component ID Description System Code Class Category Type HV-052-108 Loop B Core Spray Discharge Check Valve CS 1

A/C SA HV-051-1F015B Loop B RHR Shutdown Cooling Injection Valve RHR 1

A MO HV-051-1F017B Loop B RHR LPCI Injection Valve RHR 1

A MO

2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda.

3. Applicable Code Requirement

ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, states "Category A valves with a leakage requirement not based on an Owner's 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages are within acceptable limits. Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied."

ISTC-3630(a), Frequency, states Tests shall be conducted at least once every 2 yr."

By letter dated November 5, 2020, (Agencywide Documents Access and Management System (ADAMS) Ascension No. ML20280A757), the NRC re-approved Limerick Generating Station (LGS) Relief Request GVRR-8 dated December 17, 2018. This authorized LGS to implement a performance-based testing program for the pressure isolation valves (PIVs),

including those listed above. GVRR-8 allows the test interval to be extended to not exceed 3 refueling outages or 75 months.

As part of Relief Request GVRR-8, for those valves that are also 10 CFR 50 Appendix J (Appendix J) leak tested, a conservative control was established such that if any valve fails either the Appendix J or PIV Inservice Test (IST) program leakage test, the test interval for both tests would be reset to 2 years to be consistent with Appendix J, Option B requirement.

This test interval would be in effect until good performance is reestablished.

4. Reason for Request

HV-052-108 is part of the Appendix J program and it failed the PIV IST leakage test in the Spring 2024 refueling outage. HV-051-1F015B and HV-051-1F017B are also part of the Appendix J program and the as-found Appendix J, Type C leakage for these valves was above their administrative limit in the Spring 2024 refueling outage. Therefore, the test intervals for all three valves was reset to two years.

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 2 of 6 Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(2), an alternative to the requirement of ASME OM Code ISTC-3630 and approved LGS Relief Request GVRR-8 is proposed. The basis of the alternative is that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

LGS will be installing the Digital Modernization Project (DMP) in the Spring 2026 refueling outage. This DMP will replace the existing analog control logic hardware of the Reactor Protection System (RPS) instrumentation, Nuclear Steam Supply Shutoff System (NSSSS) instrumentation, the Emergency Core Cooling System (ECCS) instrumentation, the Reactor Core Isolation Cooling (RCIC) System instrumentation, and the End-of-Cycle Recirculation Pump Trip (EOC-RPT) instrumentation with a new single digital control system.

Due to the complex nature of the DMP installation, the current Spring 2026 refueling outage plan intentionally excludes work on the B train of ECCS. The B train of equipment will have controls set up at alternate locations in the plant, allowing the operators to operate this equipment outside of the Main Control Room (MCR). This is being done to maintain Shutdown Cooling and alternate Reactor Coolant System (RCS) makeup injection sources. The PIVs that CEG is requesting deferral of are all located on the B train of equipment. Performance of the IST leakage test on the requested valves introduces significant additional measures that impact nuclear and radiological safety risks.

The LGS DMP is structured around key technical and regulatory advances that have come to fruition in recent years. This structure demonstrates that large scale modernization is a viable economic and technical alternative. Given the scale of the LGS DMP, the public-facing research on which it is founded, and the demonstration of the multiple industry and regulatory initiatives, this project has significant benefits for the industry and by extension the public in the form of demonstrating that modernization can be achieved efficiently and will preserve reliable and carbon-free generation. The successful installation strategy being employed with the support of this proposed Alternative Request will facilitate a safe and error-free installation in a timely manner for such a large and significant undertaking.

5. Proposed Alternative and Basis for Use

The proposed alternative pursuant to 10 CFR 50.55a(z)(2) is a one-time exception to GVRR-8 to defer resetting the leakage test interval for the 3 PIVs to allow the next IST program leakage tests to be completed no later than completion of the Spring 2028 refueling outage.

If B ECCS PIV IST leakage testing is required to be performed in the Spring 2026 refueling outage, it would introduce significant additional measures that carry nuclear and radiological safety risks. There are two different options if these tests are to be performed:

The first option is to shut down the plant on the A train of RHR Shutdown Cooling using normal controls from the MCR. At approximately 55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months into the outage, the station would then perform a swap from the normal controls on the A shutdown cooling train to

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 3 of 6 the alternate controls for this same train which will need to be installed at the Remote Shutdown Panels. This is being done to maintain outage shutdown safety defense-in-depth and would require another Shutdown Cooling system swap mid outage. This would require four additional Motor Operated Valve (MOV) test boxes and three more 4kV Breaker test boxes. Three additional system indicator hook-ups in the field would be required, along with new temporary Residual Heat Removal (RHR) Heat Exchanger controller wiring. This approach also introduces additional unplanned terminations and complicates B ECCS system refill under interim operating conditions due to additional alternate controls to support the fill and vent. This additional work to prepare the A train would also incur added dose to the workers. It also necessitates a complex and infrequently performed evolutionswapping shutdown cooling trains from A ECCS (on alternate controls) to B ECCS (also on alternate controls). This would require multiple teams in multiple locations in different enclosures with complicated communications. While operators are trained for this task, the dual use of alternate controls increases the risk profile unnecessarily.

Alternatively, a full core offload could be pursued, requiring the removal of every fuel bundle from the reactor core to the spent fuel pool. This introduces reactivity management challenges due to the additional fuel moves, which is not justified given the original outage scope.

In either case, 500mR of accumulated dose is expected to perform the additional work and applicable leakage testing (Appendix J, Type C and PIV IST).

In summary, performing B ECCS PIV IST Leakage tests during the Spring 2026 refueling outage introduces avoidable risks and operational complexity, and the safest path forward is to adhere to the original plan and justify an interval extension for the B ECCS PIV IST leakage tests from this outage.

Additional basis for this Alternative Request is described below for each individual valve:

Penetration X-16B: Loop B Core Spray Discharge (HV-052-108)

HV-052-108 is a 12 Anchor Darling power assisted swing check valve. PIV IST leakage testing is required in Spring 2026 refueling outage per approved Relief Request GVRR-8 since the valve failed the as-found PIV IST leakage test in 2024. This valve also failed its freedom of movement test in the 2024 refueling outage. Based on the as-found in-body inspection of the valve, the failure is suspected to be a combination of the valve potentially not traveling to the seat under test conditions (slow pressure increase) along with normal in-body wear. The valve received maintenance to correct the freedom of movement issue, and the in-body seat and disc were reconditioned with a satisfactory as-left PIV IST leakage test.

Extension of PIV IST leakage testing is requested until the Spring 2028 refueling outage due to the valves strong PIV IST leakage history before the 2024 failure (Table 1), recent maintenance, and satisfactory as-left leakage results. The HV-052-108 valve has performed reliably for over 16 years without requiring corrective maintenance until 2024 outage. This valve has not had recent maintenance, with subsequent PIV IST leakage test history, due to its satisfactory performance for over 14 years. LGS utilizes robust maintenance

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 4 of 6 procedures and practices performed by qualified technicians to perform valve repairs.

From 2008 until 2024, PIV IST leakage rates were consistently zero, with only one failure in 2024. Based on the refurbishment and reconstitution of the disc and seat and its stable historical performance, the valve is expected to have reliable leakage performance through the 2028 outage.

Additionally, the inboard PIV barriers in this penetration are the Core Spray inboard testable check valve, HV-052-1F006B with bypass valve HV-052-1F039B. The HV-052-1F006B and HV-052-1F039B valves have had satisfactory PIV IST leak test performance for the last 16 years.

The HV-052-108 will also be exercise tested in the Spring 2026 refueling outage to demonstrate functionality and freedom of movement.

Penetration X-13B: Loop B RHR Shutdown Cooling Return (HV-051-1F015B)

HV-051-1F015B is a 12 Anchor Darling Motor Operated Globe Valve. PIV IST leakage testing is required in Spring 2026 refueling outage per approved Relief Request GVRR-8 since this valves Appendix J, Type C leakage test exceeded its administrative limit in 2024.

Extension of PIV IST leakage testing is requested until the Spring 2028 refueling outage due to the valves historically satisfactory PIV IST leakage test performance (Table 1). The HV-051-1F015B valve has operated reliably for over 16 years without requiring corrective maintenance. This valve has not failed the PIV IST leakage test in 16 years with the most recent test result in 2024 of 0.0 gpm. Since 2008, measurements have indicated that PIV IST leakage rates were consistently less than 0.17 gpm. Given this stable history of acceptable PIV IST leakage performance, the valve is expected to have reliable leakage performance through the 2028 outage.

Additionally, the inboard PIV barriers in this penetration are the Shutdown Cooling inboard testable check valve, HV-051-1F050B with bypass valve HV-051-151B and RHR Differential Pressure line valve 051-1200B. These inboard valves have had satisfactory PIV IST leak test performance for the last 16 years.

HV-051-1F015B is an Active MOV and therefore tested per the ASME OM Appendix III Program. This valve is a Low Risk, High Margin valve that is diagnostically tested. The last diagnostic test was performed in the 2024 refueling outage. There were no anomalies identified during this test that would adversely impact valve functionality, and the results ensure reliable performance and consistent seating force for acceptable leakage performance through 2028. HV-051-1F015B will be exercised and timed in Spring 2026 refueling outage to demonstrate functionality and ability to meet the required isolation time.

Penetration X-45B: Loop B RHR Low Pressure Coolant Injection (LPCI) (HV-051-1F017B)

HV-051-1F017B is a 12 Anchor Darling Motor Operated Gate Valve. PIV IST leakage testing is required in Spring 2026 refueling outage per approved Relief Request GVRR-8 since the valve exceeded its administrative limit and subsequently received maintenance in 2024.This valve exhibited high Appendix J, Type C leakage in 2024 refueling outage and

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 5 of 6 in-body maintenance was performed. Based on the as-found in-body inspection of the valve, the high leakage was attributed to indications on the seat and normal wear. The in-body seat and disc were reconditioned removing the indication and the valve had satisfactory as-left Appendix J and PIV IST leakage test results. The as-left PIV IST leakage following maintenance was 0.01 GPM.

Extension of PIV IST leakage testing is requested until the Spring 2028 refueling outage due to the valves historically satisfactory performance (Table 1). This valve has not failed a PIV IST leakage test in 16 years with the most recent test result in 2024 of 0.08 gpm. In 2012, this valve had satisfactory PIV IST leakage test results of 0.06 gpm, however, the Appendix J leakage was high prompting inbody maintenance. The as-left PIV IST leakage test result was 0.03 gpm with results up to, and including, 2024 consistently less than 0.08 gpm. This historical performance demonstrated LGSs robust maintenance procedures and practices and provides confidence in the valves leakage integrity. Given the recent stable history of acceptable PIV IST leakage performance following 2012 maintenance, the valve is expected to have reliable leakage performance through the 2028 refueling outage.

Additionally, the inboard PIV barriers in this penetration are the Shutdown Cooling inboard testable check valve, HV-051-1F041B with bypass valve HV-051-142B. These inboard valves have had satisfactory PIV IST leak test performance for the last 16 years.

HV-051-1F017B is an Active MOV and therefore tested per the ASME OM Appendix III Program. This valve is a Medium Risk, High Margin valve that is diagnostically tested.

The last diagnostic test was performed following the inbody repair in 2024. There were no anomalies identified during this test that would adversely impact valve functionality, and the results ensure reliable performance and consistent seating force for acceptable leakage performance through 2028. HV-051-1F017B will be exercised and timed in the Spring 2026 refueling outage to demonstrate functionality and ability to meet the required isolation time.

6. Duration of Proposed Alternative

This Alternative Request, upon approval, will be in effect until the end of the LGS Spring 2028 refueling outage, and once completed, LGS Relief Request GVRR-8 will again be in effect for all applicable valves.

7. Precedent None.

8. References None.

CONSTELLATION GENERATION COMPANY LLC IST PROGRAM - ALTERNATIVE REQUEST Limerick Generating Station, Unit 1 10 CFR 50.55a Alternative Request GVRR-12 Page 6 of 6 Table 1: PIV IST Leakage Test Performance HV-052-108 HV-051-1F015B HV-051-1F017B Spring 2024 AF: Off-scale*

AL: 0.01 gpm 0.0 gpm AF: 0.08 gpm AL: 0.01 gpm Spring 2022 0.0 gpm 0.17 gpm 0.0 gpm Spring 2020 Note 1 Note 1 Note 1 Spring 2018 Note 1 Note 1 Note 1 Spring 2016 0.0 gpm 0.1 gpm 0.0 gpm Spring 2014 0.0 gpm 0.0 gpm 0.0 gpm

Unable to establish test pressure during the PIV IST leakage test, therefore assigned off-scale as leakage.

Note 1: Valve testing interval extended due to acceptable performance.

v t e Letter Sequence Other
EPID:L-2025-LLA-0126, (Open)
Initiation Acceptance Administration Questions Answers Results Approval... Other:
MONTHYEAR2025-12-15 [Table View]

LG-25-132, License Amendment and Alternative Request Related to One-Time Exception of Primary Containment Isolation Valve Testing

Text

Subject:

Reference:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

3. Applicable Code Requirement

4. Reason for Request

5. Proposed Alternative and Basis for Use

6. Duration of Proposed Alternative

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LG-25-132, License Amendment and Alternative Request Related to One-Time Exception of Primary Containment Isolation Valve Testing

Text

Subject:

Reference:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

3. Applicable Code Requirement

4. Reason for Request

5. Proposed Alternative and Basis for Use

6. Duration of Proposed Alternative

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