ML25104A095

Issuance of Amendment No. 239 Revise Technical Specification Actions for Reactor Coolant System Safety/Relief Valves

ML25104A095
Person / Time
Site:	Hope Creek
Issue date:	08/05/2025
From:	Marshall M NRC/NRR/DORL/LPL1
To:	Carr E Public Service Enterprise Group
References
EPID L-2024-LLA-0089
Download: ML25104A095 (1)

Text

August 5, 2025 Mr. Eric S. Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT NO. 239 RE: REVISE TECHNICAL SPECIFICATION ACTIONS FOR REACTOR COOLANT SYSTEM SAFETY RELIEF VALVES (EPID L-2024-LLA-0089)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 239 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station. This amendment consists of changes to the technical specifications (TS) in response to your application dated June 28, 2024.

The amendment revises TS 3.4.3 Safety/Relief Valves to modify the code safety valve function lift settings for all 14 valves to 1,130 pounds per square inch gauge (psig) as well as, expand the as-found safety function lift setpoint tolerances that are listed in surveillance requirement (SR) 3.4.3.1 from +/-33.9 psig to a range of +33.9 psig to -56.5 psig. The amendment would also, increase the inservice testing (IST) program test pressure from 1,255 psig to 1,281 psig in SR 3.1.7.7, associated with TS 3.1.7, Standby Liquid Control (SLC) System.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Michael L. Marshall, Jr., Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

1. Amendment No. 239 to Renewed License No. NPF-57

2. Safety Evaluation cc: Listserv PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No.239 Renewed License No. NPF-57

1.

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

A.

The application for amendment filed by PSEG Nuclear LLC dated June 28, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-57 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented during the Fall 2025 refueling outage.

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

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 5, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.08.05 14:05:17 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 239 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Renewed Facility Operating License No. NPF-57 with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.1.7-3 3.1.7-3 3.4.3-1 3.4.3-1

reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3902 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-57 Amendment No. 239

SLC System 3.1.7 Hope Creek 3.1.7-3 Amendment 239 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each SLC subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

In accordance with the Surveillance Frequency Control Program SR 3.1.7.7 Verify each pump develops a flow rate 41.2 gpm at a discharge pressure 1281 psig.

In accordance with the INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from pump into reactor pressure vessel.

In accordance with the Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between storage tank and pump suction is unblocked.

In accordance with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored to 70°F

S/RVs 3.4.3 Hope Creek 3.4.3-1 Amendment 239 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)

LCO 3.4.3 The safety function of 13 S/RVs shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required S/RVs inoperable.

A.1 Be in MODE 3.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the required S/RVs are as follows:

Number of Setpoint S/RVs (psig) 14 1130 + 33.9

- 56.5 Following testing, lift settings shall be within +/- 1%.

In accordance with the Surveillance Frequency Control Program

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated June 28, 2024 (Agencywide Documents Access and Management System, Accession No. ML24180A127), PSEG Nuclear LLC (PSEG, or the licensee) applied for a license amendment to Facility Operating License No. NPR-57 for the Hope Creek Generating Station (Hope Creek). The licensee requested to modify technical specifications (TS) to:

(1) revise the code safety valve function lift settings to 1,130 pounds per square inch gauge (psig) for all 14 safety relief valves (SRVs) listed in TS 3/4.4.2, (2) expand the as found-safety function lift setpoint tolerances that are listed in TS 3/4.4.2 from +/-3 percent to a range from of +3 percent to -5 percent, and (3) increase the inservice testing (IST) program test pressure from 1,255 psig to 1,281 psig in Surveillance Requirement (SR) 4.1.5.c associated with TS 3/4.1.5, standby liquid control (SLC) system.

1.1

System Description

There are 14 SRVs at Hope Creek which are located on the main steam lines between the reactor vessel and the first main steam isolation valve within the drywell. Each SRV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The SRVs have two modes of operation: the spring mode and the power actuation mode.

In the spring mode of operation, the spring-loaded-pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve.

In the power actuation mode of operation, pneumatic pressure acting on a mechanical linkage assembly is used to open the valve, initiated by switches located in the control room or by pressure-sensing instrumentation. SRVs that are used in the power actuation mode are used as part of the low-low set (LLS) and automatic depressurization system (ADS) functions. The proposed setpoint change is for the spring mode of operation and does not change the opening pressure for either the LLS or ADS functions.

One of the two main functions of the SRVs is to prevent the reactor coolant system from being pressurized above the safety limit of in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code. The other main function of the SRVs is their use in the ADS where the valves open automatically as part of the emergency core cooling system (ECCS) for events involving small breaks in the reactor coolant pressure boundary (RCPB).

As stated in the Enclosure to the license amendment request (LAR), the SRVs originally installed at Hope Creek were Target Rock (TR) Model 7567F, 2-stage safety relief valves.

Similar to other industry operating experiences, the TR 2-stage SRVs at Hope Creek suffered chronic setpoint drift due to corrosion bonding. To remedy this and to ensure adherence to the

+/-3 percent end of the lift setpoint tolerance required by TS, Hope Creek replaced the original TR 2-stage SRVs with TR 3-stage Model 0867F SRVs over the course of several operating cycles.

The SLC system is a backup that operates if not enough control rods are inserted into the reactor core to accomplish shutdown and cooldown in the normal manner. The SLC system contains two positive displacement injection pumps sized to inject boron solution into the reactor at the nominal rate of 43 gallons per minute (gpm). The SLC system is designed to automatically initiate upon receipt of a signal from the redundant reactivity control system logic.

Low vessel water level, high reactor vessel dome pressure, or manual actuation starts a timer. If the core power is not downscale as indicated by the average power range monitors at the end of the time delay, SLC system operation is initiated.

1.2 Proposed Changes As stated in the Enclosure to the LAR, the licensee and industry operating experience of similar boiling-water reactors (BWRs) is that the TR 3-stage SRV models are more prone to and less tolerant of first stage pilot leakage than the previous TR 2-stage models. Hope Creek has experienced four reactor shutdowns in the past few years associated with TR 3-stage SRV leakage. Industry investigation of the leakage-related events with the TR 3-stage models supports that their reliability is dependent on several plant specific parameters including:

(1) simmer margin, which is defined as the difference in pressure between rated reactor pressure and the SRV lift setpoint, (2) maintenance practices that affect tolerances, and (3) steam line velocities which can produce undesirable acoustic and dynamic pressures near or within these valves. Therefore, in order to decrease the amount of leakage, the licensee is proposing to increase the simmer margin by changing the two lower tier SRV nominal mechanical relief setpoints (1,108 psig and 1,120 psig) in limiting condition for operation (LCO) 3.4.2.1 to 1,130 psig for all SRVs.

Given the proposed increase in setpoints for the two lower SRV setpoint groups, the overall system pressure will increase during overpressure events. To ensure that testing is consistent with the SLC system design and licensing basis, the licensee proposed changes to the surveillance criteria in SR 4.1.5.c. The current 41.2 gpm at greater than or equal to 1,255 psig is proposed to be changed to 41.2 gpm at greater than or equal to 1,281 psig. This will ensure that the SLC system can inject at higher pressures that may occur due to the increased SRV setpoints.

As stated in the Enclosure to the LAR, industry operating experience has shown that the TR 3-stage SRVs have a history of setpoint drift in the negative (or conservative) direction. This is thought to be due to a bellows-relaxation phenomenon which over time lowers the SRV lift setpoint. The current acceptance criteria for as-found lift pressure are +/-3 percent of the set pressure. Given the lowering setpoint over time, the licensee has proposed relaxing the lower as-found tolerance from -3 percent to -5 percent. The as-found tolerances are used for determining operability and to increase sample sizes for testing per the ASME Operation and Maintenance (OM) Code. Note that the proposed change is only applicable to the as-found tolerance, as the as-left tolerance will remain at +/-1 percent of the setpoint.

2.0 REGULATORY EVALUATION

The regulatory requirements for the content required in the TSs are provided Title 10 of the Code of Federal Regulations (10 CFR) 50.36, Technical specifications. The requirements in10 CFR 50.36(b) states, in part, that TS will be derived from the analyses and evaluation included in the safety analysis report.

Section 50.36(c)(2) of 10 CFR requires that TS include LCOs and states, in part, that such TS are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Further, [w]hen a LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the [TS] until the condition can be met. The regulation, 10 CFR 50.36(c)(2)(ii), further states that LCOs must be established for each item that meets one or more of the following four criteria:

Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2. A process variable, design feature or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Section 50.36(c)(3) of 10 CFR states: [SR] are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Accordingly, pursuant to 10 CFR 50.36(c)(2)(ii)(C), the SRVs fall under Criterion 3 because they are part of the primary success path and are accounted for in the Hope Creek Updated Final Safety Analysis Report (UFSAR) in the accident and safety analyses to mitigate the effects of the design basis transients and accidents. Since the setpoints in the TS for the SRVs are proposed to be modified, the licensee is required to provide acceptable analyses to support the adequacy of the TS changes.

General Design Criterion (GDC) 15, Reactor Coolant System Design, of Appendix A to 10 CFR Part 50 states [t]he reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

The ASME BPV Code, as incorporated in 10 CFR 50.55a, Codes and standards, requires that each vessel designed to meet Section III be protected from overpressure. The BPV Code allows a peak allowable pressure of 110 percent of vessel design pressure. As stated in Section 5.2.2.1.3 of the Hope Creek UFSAR, the reactor pressure vessel design pressure is 1,250 psig. This results in a peak allowable pressure of 1,375 psig. In addition, the safety limit in TS 2.1.3, Reactor Coolant System Pressure states that reactor vessel steam dome pressure shall not exceed 1,325 psig.

The ASME Operation and Maintenance of Nuclear Power Plants (OM Code), as incorporated in 10 CFR 50.55a, requires IST of pumps and valves that are within the scope of the ASME OM Code, including the SRVs within the scope of this LAR and the motor-operated valves discussed later in this Section.

Regulations in 10 CFR 50.62, Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants, provides requirements for boiling water reactor SLC system. Specifically, 10 CFR 50.62(c)(4) states that:

Each boiling water reactor must have a standby liquid control system (SLCS) with the capability of injecting into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the resulting reactivity control is at least equivalent to that resulting from injection of 86 gallons per minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design.

Therefore, any changes to the SLC system surveillance requirements require the licensee to provide acceptable analyses to support the adequacy of proposed TS change.

Based on past operating and research experience, the U.S. Nuclear Regulatory Commission (NRC or the Commission) initiated a regulatory action in the 1980s regarding testing, inspection, and maintenance of motor-operated valves (MOVs) in nuclear power plants to provide assurance that they will function when subjected to design basis conditions during both normal operation and abnormal events within the design basis of the plant. Pursuant to 10 CFR 50.55a(b)(3)(ii), the NRC conducted rulemaking, based on the results of the MOV testing programs, to require as a compliance backfit that the licensees must establish a program to ensure that MOVs continue to be capable of performing their design basis safety functions in addition to the ASME OM Code testing requirements. The licensee needed to verify the acceptable performance of MOVs in light of the SRV adjustments at Hope Creek requested in this LAR.

3.0 TECHNICAL EVALUATION

In determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the regulations, guidance, and licensing and design basis information discussed in this safety evaluation (SE). The NRC staff reviewed the licensees statements in the LAR (and its attachments and selected referenced documents), the relevant sections of the licensees UFSAR to determine if the proposed changes are acceptable.

When Hope Creek was initially licensed, the allowable as-found tolerance for the SRVs was

+/-1 percent and was consistent with all other BWRs at the time. However, it was determined that the SRVs had trouble meeting such a tight tolerance around the setpoint. In the late 1980s and early 1990s, GE Nuclear Energy developed topical report NEDC-31753P, BWROG [Boiling Water Reactor Owners Group] In-Service Pressure Relief Technical Specification Revision Licensing Topical Report (non-publicly available) to support changes to the TS requirements for the spring safety and SRVs for General Electric BWRs. Specifically, the report supported relaxation of the current Inservice opening pressure setpoint tolerance in the TS for the SRVs from +/-1 percent to +/-3 percent. This topical report was approved by the NRC on December 9, 1992 (ML20126E038). The licensee stated in the Enclosure to the LAR that current proposed technical specification changes were evaluated using the previously accepted methodology of NEDC-31753P, and its associated SE. The licensee conclusions are based on analyses and evaluations performed by both GE Hitachi Nuclear Energy (GEH) and PSEG.

Therefore, the NRC staff concludes that methodology of NEDC-31753P is acceptable as the approved methodology used by the licensee addresses, as it the issues associated with a change to the SRV lift pressures.

The main objective of the nuclear pressure relief system is to prevent overpressurization of the nuclear system. To meet this objective, the licensee must demonstrate that the peak pressure remains below acceptable limits. The licensees proposed change to increase the two lower tier SRV setpoints from 1,108 psig and 1,120 psig to 1,130 psig will increase the peak pressure during overpressure events. In addition to vessel overpressure, there are several other items that are potentially affected by this change, such as: high pressure makeup systems, containment response and dynamic loading, and ATWS response. These items and other topics are discussed below as follows: Overpressure protection in section 3.1, high pressure system performance in section 3.2, other technical considerations in section 3.3, simmer margin in section 3.4 and surveillance test history in section 3.5.

3.1 Overpressure Considerations As stated in Section 3.2.1 of the Enclosure to the LAR, there are three pressure limits in effect at Hope Creek which are unchanged with respect to the proposed changes. The first is the TS safety limit of 1,325 psig, which is measured in the vessel steam dome. The second is the ASME BPV Code limit of 110 percent of vessel design pressure or 1,375 psig. The third limit is the ASME BPV Code Service Level C limit applicable to infrequent events such as ATWS. For all the overpressure evaluations performed as part of this LAR, a single SRV is assumed out-of-service consistent with TS LCO 3.4.2.1. The remaining 13 SRVs are conservatively assumed to open at the maximum allowed pressure of 1,164 psig (1,130 psig +3 percent allowance for setpoint drift).

The licensees evaluations determined that the limiting high pressure transient event is the main steam isolation valve (MSIV) closure with flux scram. For this event, a failure of the MSIV position scram signal is assumed, and the reactor instead scrams on a subsequent high neutron flux signal due to the collapse of voids following vessel pressurization. The analysis was performed using the approved Transient Reactor Analysis Code-General Electric (TRACG) and General Electric Standard Application for Reactor Fuel methodologies. The analysis for the cycle 25 analysis resulted in a peak reactor vessel pressure of 1,289 psig, which is below the TS safety limit of 1,325 psig and ASME BPV Code limit of 1,375 psig. The licensee also evaluated the overpressure as part of a separate LAR to transition Hope Creek to a 24-month fuel cycle. This evaluation considered both 18-and 24-month fuel cycles. In all cases, the licensee identified that the peak vessel pressure was below both the TS safety limit and ASME BPV Code limit. Therefore, the NRC staff concludes that the proposed increase in SRV setpoint is acceptable because the analysis demonstrates that the overpressure criteria will not be exceeded.

Further, for ATWS, in addition to the ASME BPV Code Service Level C limit of < 1,500 psig at the vessel bottom, several other acceptance criteria were evaluated including:

peak cladding temperature is within the 10 CFR 50.46 limit of 2,200 degrees Fahrenheit (°F),

peak cladding oxidation within the 10 CFR 50.46 limit of < 17 percent, peak local suppression pool temperature less than the design limit of 217.5 °F, and peak containment pressure less than the design limit of 62 psig.

As with the overpressure evaluation, the ATWS analysis assumes 1 SRV out-of-service with the remaining 13 having a setpoint of 1,130 psig and an allowable setpoint drift of +3 percent.

Two ATWS scenarios were evaluated, including MSIV closure with no scram, and a pressure regulator failed open with no scram. The limiting pressure was found in the pressure regulator failed open case with a peak pressure of 1,473 psig, which is below the acceptance criteria of 1,500 psig. The other acceptance criteria (items 1 through 4 listed above) are not significantly affected by the change in SRV setpoint. The results provided in Section 7 of Attachment 3 (public available) and 5 (non-public available) of the LAR show that all acceptance criteria have been met; therefore, the NRC staff concludes that the proposed increase in SRV setpoint is acceptable because the ATWS analysis demonstrates that the overpressure criteria will not be exceeded.

3.2 High pressure system performance Three high pressure systems were evaluated by the licensee to assess any effects given the proposed changes. These include the high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and SLC systems which are discussed individually below. The licensee also performed evaluations to assess the impact of the SRV setpoint pressure increase to 1,130 psig on valves associated with the high-pressure systems as well as other valves (i.e., main steam) that may be impacted by an increase in reactor pressure. According to the licensee, the assessment identified that no physical changes are required associated with HPCI, RCIC, or other MOVs.

3.2.1 High-pressure coolant injection As described in Hope Creek UFSAR Section 6.3.1.2.1, the primary purpose of HPCI is to maintain reactor vessel inventory after small breaks that do not depressurize the reactor vessel.

The HPCI system is also used to maintain reactor vessel inventory following reactor isolation and coincident failure of the non-ECCS RCIC System. The system pumps water through one of the core spray spargers and one of the feedwater spargers.

An increase in the SRV opening pressure will only affect events in which there will be SRV actuations. For large break sizes, reactor pressure decreases from the point in which the break initiates, therefore, the proposed SRV setpoint change has no effect on HPCI. For smaller breaks, pressure will initially decrease, however, it increases and peaks after the MSIVs close.

The HPCI system is flow-controlled to provide 5,600 gpm. An increase in system pressure would be compensated for by the HPCI speed control system, which would increase turbine speed to overcome the increase in pressure. The licensee states that margins in existing design calculations support that HPCI is fully capable of supplying its design required 5,600 gpm with all SRVs at the 1,130 psig setpoint with a +3 percent setpoint drift tolerance without any change to the current turbine operating speed limit of 4,150 rpm or overspeed trip setpoint. At the design operating speed, HPCI currently has 148 feet (ft) of total discharge head (TDH) margin.

The licensee identified that this margin exceeds the 53 ft of additional TDH needed to accommodate the SRV setpoint increase. Operation of HPCI up to 4,225 revolutions per minute has been evaluated as satisfactory if additional margin is needed. In addition, the licensee conservatively did not take credit for the LLS function in the evaluations. The LLS function would result in reduced pressures and additional margin. Therefore, based upon its review of the aforementioned information, the NRC staff concludes that there is reasonable assurance that the HPCI will continue to perform its safety function with the proposed increase to the SRV setpoint because HPCI will be able to maintain reactor vessel inventory.

3.2.2 Reactor Core Isolation Cooling System As described in Hope Creek UFSAR Section 5.4.6.1, the RCIC system is a safety-related system designed to ensure that sufficient reactor water inventory is maintained in the reactor vessel to allow for adequate core cooling. Similar to the HPCI system, RCIC is flow-controlled to provide its rated flow of 600 gpm. An increase in system pressure would be compensated for by the RCIC speed control system which would increase turbine speed to overcome the increase in pressure. Per the design calculations, at the design operating speed, RCIC currently has 251 ft of TDH margin which exceeds the 53 ft of additional TDH needed to accommodate the increase in SRV setpoint. In addition, no credit for the LLS function is taken for these evaluations which would add additional margin. Therefore, based upon its review of the aforementioned information, the NRC staff concludes that there is reasonable assurance that the RCIC will continue to perform its safety function with the proposed increase to the SRV setpoint because RCIC will be able to maintain reactor vessel inventory.

3.2.3 Standby Liquid Control System In Hope Creek UFSAR Section 9.3.5.1, it states that means are provided by which the functional performance capability of the system components can be verified periodically under conditions approaching actual use requirements. TS SR 4.1.5.c specifies the flow rate and discharge pressure necessary for each SLC pump to meet the requirements of 10 CFR 50.62(c)(4). In order to meet these requirements with the proposed increase in SRV setpoint to 1,130 psig, the licensee is proposing to increase the required pump discharge pressure from 1,255 psig to 1,281 psig.

As discussed in Section 9.3.5 of the Hope Creek UFSAR, the SLC system is a backup for reactivity control independent of normal reactivity control provisions in the nuclear reactor, to shut down the reactor if the normal reactivity control provisions become inoperative. The SLC pumps are of the positive displacement type and provide a constant flowrate regardless of system pressure up to the design pressure of 1,400 psig. The SLC pumps are not assumed to be operating until 230 seconds after an ATWS event, at which point, pressures are well below 1,400 psig. The licensee asserts that the proposed change to the SRV setpoints does not result in any adverse impact to the SLC system. Additionally, the licensees evaluation determined that neither the boron shutdown concentration, nor the injection rate requirement requires change due to the effects of the proposed increased SRV setpoints. Assuming the nominal pump flowrate, the licensee determined the required discharge head is 1,281 psig. This value includes a reactor pressure of 1,177.3 psig plus 103 pounds per square inch differential (psid) of system elevation and head loss. Therefore, the NRC staff, based upon its review of the aforementioned information, concludes that there is reasonable assurance that the SLC system will continue to perform its safety functions with the proposed increase to the SRV setpoint because the SLC system's design and operational parameters have be analyzed with the increased pressure and did not present a compromise to its safety functions. Further, NRC staff also concludes that the change to SR 4.1.5.c to increase the pressure to 1,281 psig is acceptable because the proposed change is consistent with the expected conditions when the SLC pumps are in operation.

3.3 Other Technical Considerations 3.3.1 Loss-of-Coolant Accident (LOCA)

As discussed above in this SE, the proposed increase in SRV setpoint to 1,130 psig has no effect on a large break LOCA as the system rapidly depressurizes upon break initiation so the SRV would not be needed. For smaller breaks, the SRVs may open once the MSIVs close and pressure increases, however, these breaks are not limiting, so the proposed increase to the SRV setpoint does not change the limiting LOCA evaluations. In addition, for the smaller break sizes, the ADS is credited in the safety analysis to open the selected SRVs to reduce system pressure, and the ADS setpoints are not reliant on the SRV setpoints. Therefore, based upon its review of the aforementioned information, the NRC staff concludes that the LOCA analysis are not affected by the proposed increase to the SRV setpoint because it does not change the limiting LOCA evaluations.

3.3.2 Containment The licensee evaluations for the containment are described in Section 6 of Attachments 3 (non-proprietary) and 5 (proprietary) to the LAR. These evaluations considered the following items: short-term containment response and hydrodynamic loads, long-term suppression pool temperature for Design Basis Accident-LOCA and net positive suction head (NPSH) analyses, long-term suppression pool temperature for loss of power and NPSH analyses, intermediate line break accident and small line break accident analyses for Plant Unique Load Definition, local suppression pool temperature for NUREG-0783 analysis, and the steam line break drywell temperature response for the evaluation of drywell environmental qualification considerations.

For events such as the limiting design basis LOCA, the vessel depressurizes without the operation of any SRVs. Therefore, NRC staff finds that the proposed changes to the SRV setpoints do not have an impact on containment pressure/temperature or NPSH response for the design basis LOCA.

For the loss of offsite power (LOOP) evaluation, the licensee evaluation shows there is no SRV operation prior to the LLS relief logic which occurs above the scram setpoint and below the lowest allowable SRV opening pressure. Therefore, NRC staff concludes that the proposed changes to the SRV setpoints, and tolerances have no impact on the LOOP event For intermediate and small steam line breaks, the existing licensee evaluation is based on an endpoint-type calculation, which is controlled by the amount of initial stored energy in the primary system and decay heat. The NRC staff finds that the existing analysis remains applicable.

For the NUREG-0783 local suppression pool temperature, the limiting event in the existing licensee evaluation is not a pressurization transient that results in SRV operation. Therefore, NRC finds that the licensees existing NUREG-0783 local suppression pool temperature evaluation is not impacted by the proposed change to the SRV setpoints because this event would not result in SRV operation.

With the proposed increase in the two lower-group SRV setpoints from 1,108 psig and 1,120 psig to 1,130 psig (with the same allowable tolerance of +/-3 percent), the SRV opening pressure may increase which would result in increased SRV flow and additional dynamic loading. These dynamic loads include reaction and thrust loads acting on the SRV discharge line and T-quencher and their supports, and air-bubble pressure loads on submerged pool boundary and air-bubble drag loads on submerged structures. To evaluate these loads, the licensee used the same methodology as previously used when they increased the setpoint tolerance from +/-1 percent to +/-3 percent. The licensee evaluation identified that the additional loads were found to be much smaller than the available conservatisms in the existing analyses.

Therefore, based upon its review of the aforementioned information, NRC concludes that the proposed change to the SRV setpoints does not negatively impact the containment analysis.

3.3.3 Fuel Per NEDC-31753P, the effect of the SRV opening pressure on the minimum critical power ratio (MCPR) response must be determined for abnormal operational occurrences. To address this, the licensee reanalyzed the cycle 25 reload pressurization transients with the increased SRV setpoint of 1,130 psig. The evaluation confirmed that the cycle 25 MCPR limits remained valid and there continued to be margin to the thermal and mechanical design limits. The cycle-specific transient reload licensing analyses will provide transient event fuel thermal limits for each fuel cycle to confirm the limits are still met. Therefore, based upon its review of the aforementioned information, NRC staff concludes that the MCPR will not be challenged given the proposed increase to the SRV setpoint because the conclusions within licensees analysis confirmed that the aforementioned limits are still met and there continues to be margin to the thermal and mechanical design limits.

3.3.4 Other items The licensee evaluated how the proposed change to the lower setpoint tolerance from -3 percent to -5 percent would impact the LLS logic and reactor protection system (RPS) actuation setpoint. The LLS logic is actuated at 1,047 psig +/-2 percent resulting in LLS operation from 1,026 psig to 1,068 psig. The RPS reactor vessel steam dome pressure-high allowable value is 1,057 psig. Given the proposed change in setpoint tolerance, the lowest actuation of an SRV is at 1,073.5 psig (1,130 psig

• 0.95). Therefore, the NRC staff concludes that. that the proposed change to the lower setpoint tolerance from -3 percent to -5 percent has no impact on the LLS or RPS high pressure trip because this value is above both the LLS and RPS actuation setpoint.

Additionally, the licensee performed an evaluation of the proposed changes on the downcomer piping, supports, spargers, containment, and suppression pool loads. All were found to be acceptable given the proposed change to the SRV setpoints.

3.4 Simmer Margin The simmer margin is defined as the difference between normal reactor operating pressure and the SRV setpoint. In the Enclosure to the LAR, the licensee stated that in GE Service Information Letter 196, Supplement 3, the manufacturer recommended simmer margin is 120 psi. Using the proposed setpoint of 1,130 psig for all SRVs results in a simmer margin of 125 psi (given the 1,005 psig nominal operating pressure). The increase in simmer margin from the proposed increase to the SRV setpoints is expected to contribute to a decreased probability of SRV pilot stage leakage.

As discussed in Section 3.1 of the Enclosure to the LAR, the licensee states that the TR 3-stage SRV pilot stem begins to move at a pressure lower than its lift setpoint. This pressure, referred to as the abutment pressure, is 92 percent of the SRV setpoint. Taking the current lowest setpoint of 1,108 psig and accounting for the abutment pressure, the pilot stem could begin moving at 1,019.4 psig (assuming no setpoint drift). Taking the proposed increase in SRV setpoint to 1,130 psig, the pilot stem would begin moving at 1039.6 psig which results in an improvement to the simmer margin of 20.2 psid (and 9.2 psid for the SRVs with a setpoint at 1,120 psig).

Repeating the above where the allowable setpoint drift is taken into account, the SRVs with the 1,108 psig setpoint and a -3 percent drift could result in the pilot stem movement beginning at 988.8 psig. With the proposed 1,130 psig setpoint and -5 percent allowable drift, the pilot stem could begin movement at 987.6 psig, which is practically unchanged over the current configuration.

Therefore, based upon its review of the aforementioned information, NRC staff concludes that the changes to increase the SRV setpoint to 1,130 psig with an allowable as-found tolerance of

+3 percent/-5 percent will not negatively impact the simmer margin because the SRVs are expected to contribute to a decrease probability of SRV pilot stage leakage.

3.5 Surveillance Test History As stated above in this SE, the licensee replaced the original TR 2-stage SRVs with TR 3-stage SRVs over the course of several refueling outages. As part of the Enclosure to the LAR, the licensee provided the results of the surveillances from the 3-stage SRVs. Overall, the Hope Creek 3-stage SRV test results show an average drift of -1.42 percent. Out of the tested SRVs, only one failed the surveillance. This SRV had as -found setpoint of 1,085 psig with a nominal setpoint of 1,130 psig, resulting in a drift of -3.98 percent, which is outside the +/-3 percent allowable value. The licensee also stated that their operating experience with the TR 3-stage SRVs is consistent with those at the James A FitzPatrick Nuclear Power Plant whose SRVs include testing up to a span of four years between tests versus the three years maximum at Hope Creek.

The licensee proposed lowering the as-found lower setpoint tolerance from -3 percent to -5 percent to avoid unnecessarily declaring the SRVs inoperable and requiring expanded scope testing, in accordance with the requirements of the ASME OM Code. Inservice testing requirements, as required by the ASME OM Code as incorporated in 10 CFR 50.55a, will continue to be met through implementation of the licensees IST Program and the licensees compliance with these separate regulatory requirements are unaffected by the requested changes in the LAR. Since the safety function of the SRVs is to prevent overpressure conditions, opening earlier (i.e., below the setpoint) results in lower peak pressures. As discussed above in of this SE, the simmer margin is maintained. Therefore, based upon its review of the aforementioned information, the NRC staff finds the proposed as-found setpoint low tolerance of -5 percent acceptable because the proposed SRV changes will result in the SRV opening earlier.

4.0 TECHNICAL CONCLUSION The licensee proposed changes to TS LCO 3.4.2.1 to 1) increase the SRV setpoints from 1,108 psig and 1,120 psig to 1,130 psig and 2) decrease the lower as-found tolerance from -3 percent to -5 percent. In addition, the licensee proposed to change SR 4.1.5.c to increase the SLC test pressure from 1,255 psig to 1,281 psig. The NRC staff has reviewed the licensees analysis provided in the Enclosure to the LAR as well as GEH report NEDC-34037P, Safety Review for Hope Creek Safety/Relief Valve Setpoint Increase and Tolerance Change, Rev. 0, May 2024 (included as an attachment to the LAR) and finds that with the proposed increase to the SRV setpoint, setpoint tolerance change and SLC test pressure change that:

The reactor coolant system pressure safety limit in TS 2.1.3 will be protected, The ASME BPV Code requirement that peak allowable pressure of 110 percent vessel design pressure will not be exceeded, The ECCS/LOCA analysis are not affected, The high-pressure systems (HPCI, RCIC and SLC) will continue to perform their safety functions, Containment loadings will remain within the design basis limits, and, The ATWS acceptance criteria will not be violated.

The NRC staff concludes that the requirements of 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to assure that the necessary quality of systems and components are maintained that facility operation will be within safety limits, and that the LCOs will be met.

Based on these findings, the NRC staff concludes that there is reasonable assurance that the requirements of GDC 15 and the ASME BPV Code related to vessel overpressure will continue to be met; therefore, the NRC staff finds the proposed changes to be acceptable.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New Jersey State Official was notified of the proposed issuance of the amendment on January 23, 2025. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration as documented in the Federal Register (89 FR 71445, published on September 3, 2024), and there has been no public comment on such finding.

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

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

Principal Contributor: R. Beaton, NRR Date: August 5, 2025

ML25104A095 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/SNSB/BC NRR/DEX/EMIB/BC NAME TEdwards MMarshall KEntz DMurdock SBailey DATE 6/10/2025 6/16/2025 6/17/2025 11/25/2024 6/23/2025 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME SMehta KWilson HGonzález MMarshall DATE 6/30/2025 7/30/2025 8/5/2025 8/5/2025