ML25239B557

Issuance of Amendment No. 257 to Revise Updated Safety Analysis Report to Approve Alternate Method of Verifying Main Steam Isolation Valve Closure Reactor Protection System Response Time

ML25239B557
Person / Time
Site:	Clinton
Issue date:	09/08/2025
From:	Wetzel B NRC/NRR/DORL/LPL3
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
Wetzel B, NRR/DORL/LPL3
References
EPID L-2025-LLA-0099
Download: ML25239B557 (1)

Text

September 8, 2025 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

CLINTON POWER STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 257 TO REVISE UPDATED SAFETY ANALYSIS REPORT TO APPROVE ALTERNATE METHOD OF VERIFYING MAIN STEAM ISOLATION VALVE CLOSURE REACTOR PROTECTION SYSTEM RESPONSE TIME (EPID L-2025-LLA-0099)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 257 to Facility Operating License No. NPF-62 for Clinton Power Station, Unit No. 1 (Clinton). The amendment is in response to your application dated June 26, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25177C292).

The amendment authorizes revision to the Clinton Updated Safety Analysis Report to implement an alternate methodology of verifying Reactor Protection System Response Time for Main Steam Isolation Valve Closure as an NRC-approved methodology.

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

Sincerely,

/RA/

Beth Wetzel, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-461

Enclosures:

1. Amendment No. 257 to NPF-62

2. Safety Evaluation cc: Listserv CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-461 CLINTON POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 257 License No. NPF-62

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC (the licensee) dated June 26, 2025, as supplemented by letters dated August 13 and August 20, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, by Amendment No. 257, Facility Operating License No. NPF-62 is amended to authorize revision to the Updated Safety Analysis Report (USAR), as set forth in the application dated June 26, 2025, as supplemented by letters dated August 13 and August 20, 2025. The licensee shall update the USAR to incorporate the changes as described in the licensees application dated June 26, 2025, as supplemented by letters dated August 13 and August 20, 2025, and in the NRC staffs safety evaluation associated with this amendment, and shall submit the revised description authorized by this amendment with the next update of the USAR.

3.

This amendment is effective as of its date of issuance and shall be implemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: September 8, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.09.08 10:53:57 -04'00'

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 257 TO FACILITY OPERATING LICENSE NO. NPF-62 CONSTELLATION ENERGY GENERATION, LLC CLINTON POWER STATION, UNIT NO. 1 DOCKET NO. 50-461

1.0 INTRODUCTION

By letter dated June 26, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25177C292), as supplemented by responses to Request for Additional Information (RAI) dated August 6 2025 (ML25218A252), by letters dated August 13, 2025 (ML25225A226), and August 20, 2025 (ML25232A248), Constellation Energy Generation, LLC (CEG, the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC, the Commission) a license amendment request (LAR) for Clinton Power Station, Unit No. 1 (CPS, Clinton). The proposed amendment would authorize revision to the Clinton Updated Safety Analysis Report (USAR) to incorporate an alternate method of verifying Main Steam Isolation Valve (MSIV) Closure Reactor Protection System Response Time (RPSRT).

The supplemental letters dated August 13, 2025, and August 20, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 6, 2025 (90 FR 37894).

The NRC staff has reviewed the licensees submittals. Within the scope of its review, the staff finds the proposed USAR changes to be acceptable. The staffs evaluation is described below.

1.1 Proposed Change and Reason for Change In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, and 10 CFR 50.59, Changes, tests, and experiments, the licensee requested NRC approval of an alternate method of verifying MSIV Closure RPSRT. In the NRCs Comprehensive Engineering Team Inspection Report issued on May 8, 2025 (ML25128A110), the NRC issued a Green Non-Cited Violation (NCV) to Clinton. The violation was cited against 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion V, Instructions, Procedures, and Drawings, for failure to have procedures appropriate to the circumstance for performing response time testing of the reactor protection system main steam isolation valve closure function, an activity affecting quality. The violation stated that this failure had led to the licensee not meeting Clinton Technical Specification (TS) Surveillance Requirement (SR) 3.3.1.1.17, which measures RPS Response Time.

The CPS TS 1.1 Definition of REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME reads:

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification has been previously reviewed and approved by the NRC.

The licensee performs SR 3.3.1.1.17 for MSIV Closure by generating the trip signal via a test switch in the main control room. The test switch in the control room replaces the MSIV Closure sensor contact and associated wiring to the associated MSIV. The licensee requested approval of an alternate method of generating an open trip signal via a test switch in the main control room as an NRC-approved methodology.

2.0 REGULATORY EVALUATION

2.1 System Description and Background The licensee provided the background and the description of the RPSRT for the MSIV in, Evaluation of Proposed Changes, of its LAR. In Section 2 of Attachment 1 of the LAR, the licensee provided a reason for the proposed change. The licensee proposed to revise Section 1.8, Conformance to NRC Regulatory Guides, of the USAR, under Regulatory Guide

[RG] 1.118, Rev. 2 (June 1978), Periodic Testing of Electric Power and Protection Systems, to include a description of the verification methodology used to measure MSIV Closure RPSRT when performing TS SR 3.3.1.1.17. Additionally, LAR Attachment 3, the response to the NRC staffs RAI, and the supplemental information to the RAI response provide commitments by the licensee to monitor and trend the RPS MSIV Closure position as-found data from the prior two calibrations, and to verify there is no apparent degradation of the performance of the MSIV NAMCO EA740-50100 snap-acting limit switch that would appear to adversely affect the MSIV Closure RPS input channel response time.

2.2 Proposed Change The licensee proposed revising the Clinton USAR with regard to following RG 1.118, Rev. 2 by adding a new Project Position 4, as follows:

4)

Reference:

Paragraph C.3 - For RPS Time Response Testing for MSIV Closure, the substitute test creates an RPS trip signal by use of a test switch in the control room instead of at the sensor at the MSIV. This test switch replaces the MSIV Closure sensor contact and associated wiring from the associated MSIV to the control room. While this removes the wiring to and from the MSIV to the control room, this is negligible on overall response time due to the speed at which wiring de-energizes. Degradation of this sensor is assessed by periodic channel calibrations. Description and approval of this method is found in the Safety Evaluation of Amendment No [257].

2.3 Applicable Regulatory Requirements The regulation at 10 CFR 50.55a(h) states, in part, that the RPS for nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, must meet the requirements in Institute of Electrical and Electronics Engineers (IEEE) Std 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or the requirements in IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, or the requirements in IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995.

The regulation at 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 13, Instrumentation and control [(I&C)], states that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

The regulation at 10 CFR Part 50, Appendix A, GDC 21, Protection system reliability and testability, states that the protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. It also states that the protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

2.4 Applicable Regulatory Guidance RG 1.118, Rev. 2, Periodic Testing of Electric Power and Protection Systems, dated June 1978 (ML12216A012), describes a method acceptable to the NRC staff for complying with the NRCs regulations for ensuring that periodic testing of the electric power and protection systems provide means for checking the operational availability of each protection system input sensor during reactor operation. RG 1.118, Rev. 2 endorses IEEE Std 279-1971 and IEEE Std 338-1977, Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems. The NRC staff used this guidance to determine the adequacy of the licensees alternative periodic response time testing methodology/program for the Clinton RPS MSIV Closure position instrument channel and the related plant surveillance procedures.

Section 4, Basis, of IEEE 338-1977 states, in part:

Periodic testing of Class lE safety systems contributes, by means of the detection of failures, to the realization of desired system operational availability and calls attention to performance that is not within prescribed limits.

Testing programs shall be conducted in a logical sequence such that the overall condition of the systems under test can immediately be assessed and the need for progressing further into the testing of individual components be determined.

The program for each system shall be designed to produce objective data necessary for assessing the performance and the availability. The testing program should provide trend data and the capability to observe degradation and an indication of incipient failures. (emphasis added)

3.0 TECHNICAL EVALUATION

3.1 Overview of the licensees proposed verification methodology In its June 26, 2025, submittal, the licensee stated that it is requesting NRC review and approval of a proposed alternative response time verification methodology to be performed per procedure CPS 9431.20, RPS Main Steam Isolation Valve (MSIV) Response Time Test. The methodology verifies the response time of the RPS to an MSIV Closure signal. The licensee has proposed that rather than measure the MSIV Closure RPS Input channel response time by applying test equipment directly to the snap-acting limit switch on the MSIV, and record the response of the channel to snap-acting limit switch motion on the MSIV while recording the time that the RPS relays drop out in the control room area, the proposed method will be to substitute a temporary test switch located closer to the control room in place of the MSIV snap-acting limit switch to generate a test signal input to the RPS. This signal is generated by breaking (opening) the circuit via the temporary test switch in the Main Control Room (MCR). That is, for the purpose of MSIV Closure response time testing, the temporary test switch would replace the MSIV Closure snap-acting limit switch (NAMCO EA740-50100) sensor contact and associated wiring between the MCR and the associated MSIV. The model of the temporary test switch is a generic single pole test switch.

Approval of the proposed methodology would allow the licensee to comply with the Clinton TS RPSRT (SR 3.3.1.1.17) definition by verifying an approved component (MSIV snap-acting limit switch) with an NRC reviewed and approved methodology. To meet the definition of RPSRT, the test must be done at the channel sensor (i.e., the MSIV snap-acting limit switch) by either manually toggling the switch, closing the MSIV, or using the alternative methodology described in RPS Main Steam Isolation Valve (MSIV) Response Time Test performed per procedure CPS 9431.20.

The licensee is requesting the usage of the methodology described in RPS Main Steam Isolation Valve (MSIV) Response Time Test performed per procedure CPS 9431.20 due to the impracticality of timing events in the millisecond range at remote location in the plant and the MCR. Performing this test at the sensor would require connecting a high-speed chart at the MSIV snap-acting limit switch to the MCR at the output of the high voltage level input digital signal conditioner where it inputs to the solid-state RPS logic.

The LAR states that plant conditions during refueling outages do not always allow for MSIVs to be cycled open or closed to support testing needs. A closed MSIV would prevent the ability to toggle the snap-acting limit switch manually. Cycling the MSIVs would cause unnecessary wear and tear. Approval of this testing methodology will allow greater flexibility in scheduling and performing the required testing during an outage and prevent undue wear on the MSIVs.

Additionally, the LAR states that savings are realized by not having personnel at the MSIV snap-acting limit switch for this testing. Inboard MSIVs are in the drywell. Outboard MSIVs are in the Auxiliary Building Steam Tunnel. The dose estimates for the channel calibrations for MSL A Channel Cal 9431.10A, RPS MAIN STEAMLINE ISOLATION VALVE B21-F022A CHANNEL CALIBRATION, and MSL D Channel Cal 9431.10D, RPS MAIN STEAMLINE ISOLATION VALVE B21-F022D CHANNEL CALIBRATION, are 46 mrem and 69 mrem, respectively. This is representative of an average dose accumulated per Main Steam Line for RPS Response Time Testing. Performing this test from the control room instead of locally will save an average of 115 mrem per outage.

The LAR also states that stroking the MSIVs to measure RPSRT is not warranted if there is no mechanism by which valve position could become uncoupled from snap-acting limit switch actuation. In other words, snap-acting limit switch actuation is what indicates valve position, specifically, that MSIV position has reached the RPS Setpoint. The physical setting of the snap-acting limit switch is verified on a more frequent basis by calibration checks; therefore, stroking the valve to actuate the snap-acting limit switch is not required for response time purposes and introduces unnecessary wear to these components.

The MSIV-Closure RPS trip signal is verified by the following SRs, all of which are controlled by the Surveillance Frequency Control Program:

SR 3.3.1.1.9 (Channel Functional Test), currently performed at least every 24 months.

SR 3.3.1.1.13 (Channel Calibration), currently performed at least every 48 months.

SR 3.3.1.1.15 (Logic System Functional Test), currently performed at least every 48 months.

SR 3.3.1.1.17 (RPS Response Time verification), currently performed at least every 96 months per division for the 4 divisions at CPS.

The licensee proposed an alternative method to perform the MSIV Closure Instrument Channel Response Time Test in two parts, as follows:

A) RPS Instrument Channel downstream of the temporary test switch test.

The RPSRT (SR 3.3.1.1.17) testing for the control room portion of the instrument channel is performed for each MSIV as follows:

1. Leads are lifted in the MCR for the associated outboard MSIV snap-acting limit switch circuit.

2. A temporary test switch is installed across the terminal points from which the leads were lifted.

3. A high-speed chart recorder (ASTRO-Med Dash 10 or equivalent) is connected across those terminal points.

4. A high-speed chart recorder is connected at the output of the high-voltage level input digital signal conditioner where it inputs to the solid-state RPS logic.

5. The chart recorder is started then the test switch for the associated MSIV is opened.

6. Response time data is recorded.

7. The associated MSIV snap-acting limit switch circuits are restored to their pre-test configurations.

8. Total Response Time for the portion of the circuit downstream of the temporary test switch is determined for a given division by using the slower response time of the inboard and outboard MSIV and adding the Self-Test System response time (8.25 milliseconds) of the Logic/Load Driver. This overall value is verified to be 40 milliseconds. This is the MSIV Closure response time for the portion of the circuit downstream of the temporary test switch listed in the Operational Requirements Manual (ORM) Attachment 2, Table 13.

Figure 1 Figure Showing test for A outboard MSIV B) MSIV snap-acting limit switch portion of the Instrument Channel Response Time Test The adequacy of the response time for the MSIV snap-acting limit switch is based on actions taken during scheduled Calibration Surveillance Tests. Normally, there should be no observable delay in snap action of the MSIV snap-acting limit switch when the MSIV Valve Stem is closing from full open to the required setpoint at which the nominal switch setting is reached. If there is any observable delay, the switch is deemed to not meet its required response time.

The Channel Calibration Test (SR 3.3.1.1.13) verifies that a trip signal occurs within the TS Allowable Value of 13 percent closed (from the open position). This is verified by taking a physical measurement of stem travel distance from the full open position to the full closed position, locally at the MSIV, and comparing it to a physical measurement of stem travel distance required to open the snap-acting limit switch which drives the RPS trip signal. This test can reveal possible deterioration of the MSIV snap-acting limit switch, such as sluggish behavior (e.g., slow movement or sticking) of the MSIV snap-acting limit switch, which would preclude a trip within the TS 3.3.1.1-1 allowable limits. Moreover, the channel calibration ensures sufficient overlap in circuit diagnostics to identify degradation in the parts of the channel not actively monitored by the current SR 3.3.1.1.17 methodology, and it currently achieves this at an equal or higher frequency compared to SR 3.3.1.1.17.

3.2

NRC Staff Evaluation

Response Time Verification Tests (IEEE Std 338-1977, 6.3.4) are required to verify that the protection system response times are within the limits of the overall response time given in the safety analysis. The NRC staff notes that the IEEE standard provides allowances for such testing if the response time can be verified by functional testing, calibration testing, or both.

Specifically, the IEEE standard states that Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety system equipment is verified by functional testing calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine periodic tests.

The NRC staff notes that the proposed overlap of the tests performed within SR 3.3.1.1.17 with SR 3.3.1.1.9, SR 3.3.1.1.13, SR 3.3.1.1.15, and SR 3.3.1.1.17 provides several opportunities to verify that the MSIV snap-acting limit switches are responding in an appropriate manner.

Additionally, the NRC staff notes that an important consideration concerning substituting a temporary test switch for the MSIV snap-acting limit switch when recording the response time of the instrument channel equipment located in the control room is that the basis for performing response time tests is to be able to detect degradation within the components needed to accomplish the required safety functions within their safety analysis performance requirements.

When measuring the response of the channel equipment located in the control room, measurement and test equipment and recording equipment are used to document and enable trending of the actual time it takes for that portion of the RPS channel to respond to the stimulus of the temporary test switch. Verification of the appropriate response time of the MSIV snap-acting limit switch portion of the instrument channel is based on there being no observable delay in response of the snap-acting limit switch while conducting Calibration Surveillances, SR 3.3.1.1.13, which are performed twice as frequently as the Response Time Surveillance SR 3.3.1.1.17. This is acceptable because the IEEE standard states that the testing program should provide trend data and the capability to observe degradation and an indication of incipient failures. (emphasis added)

The NRC staffs review focused on determining whether there is reasonable assurance that:

1. Changes in the MSIV Closure snap-acting limit switch performance characteristics beyond acceptable thresholds would be reliably detectable when performing the calibration process using the frequency required for each calibration SR.

2. Instructions for carrying out the proposed methodology would provide adequate and sufficiently detailed information for the plant maintenance and engineering staff to observe, identify, and document/record actual MSIV Closure snap-acting limit switch performance degradation, if any, from one calibration surveillance to the next and when trended over time, and initiate appropriate corrective action for the snap-acting limit switch when the threshold for observed degradation has been reached.

The NRC staff evaluated the materials within the licensees LAR, its RAI response, and its supplemental information. The staff noted that the licensees proposed revised calibration procedure and supplemental information included the following new statements:

1. Performing SR 3.3.1.1.13 (channel calibration) at least every 48 months and SR 3.3.1.1.17 (RPS response time verification) at least every 96 months provides sufficient overlap to detect possible deterioration such as sluggish behavior (e.g., slow movement or sticking) of the MSIV snap-acting limit switch, which would preclude a trip within the TS 3.3.1.1-1 allowable limits. Also, the channel calibration provides an opportunity to trend the as-found setpoint of the MSIV valve stem travel stroke for snap-acting limit switch degradation. Moreover, the channel calibration ensures sufficient overlap in circuit diagnostics to identify degradation in the parts of the channel not actively monitored by the current SR 3.3.1.1.17 methodology, and it currently achieves this at an equal or higher frequency compared to SR 3.3.1.1.17.

2. The calibration procedure explanation, as provided in the supplemental information letter dated August 20, 2025, provides sufficiently detailed instructions that enable the plant maintenance and engineering staff to detect and evaluate degradation in a key safety component (e.g., valve stem, snap-acting limit switch) during calibration surveillances.

The instructions also enable the plant maintenance and engineering staff to initiate corrective action to the snap-acting limit switch performance when needed and prevent its failure before it is needed to perform its required safety function.

a. As stated in the supplemental information letter: [T]he calibration procedure requires coordination between the Main Control Room (MCR) operator controlling the valve, an Instrumentation Technician in the MCR monitoring electrical response (observation of voltage change from nominal 125 VDC to 0 VDC), and an Instrumentation Technician locally at the MSIV observing operation of the valve and limit switch and marking valve position when the limit switch electrically changes state. Instrumentation Technicians that have experience performing this test report that switch actuation is normally rapid as soon as the setpoint is reached due to the spring-loaded operation of the switch. The Instrumentation Technician in the MCR can audibly hear the switch actuation over the headphones while observing the position indicating lights and digital voltmeter. In this way, any erratic electrical response can be observed. Because the limit switches are bolted in place and the actuator arms are splined to the shaft, other than delayed response, there is no mechanism for drifting of the limit switch setpoint. Any delay in the response of the limit switch would be manifested as further valve travel when being monitored electrically. Changes in measured as-found limit switch setting can be trended to determine if any delay is occurring.

3. The proposed alternate methodology described in CPS procedures 9431.10 (A through H) for RPS Main Steamline Isolation Valve channel calibrations instructs the plant staff to monitor the snap-acting limit switch for sluggish and delayed behavior in step 8.1.15 of the procedure and use step 9.2.4 to notify supervision for failure to meet the acceptance criteria.

As outlined above, the NRC staff was able to evaluate the licensees proposed alternative approach for response time testing the MSIV Closure Instrument Channel. The staff determined that the proposed revisions to the plant procedures described in the LAR supplements to implement the alternative methodology, as outlined above, would provide reasonable assurance that effective and repeatable actions would be taken by plant personnel when measuring, trending, and observing performance of the snap-acting limit switch during a combination of channel calibration surveillances and response time surveillances. Such proposed revisions would also signal the need for plant personnel to take appropriate remedial actions to repair or replace any identified degraded equipment within the instrument channel as appropriate.

Therefore, the staff finds that implementation of the proposed alternative methodology outlined above addresses all the appropriate principles discussed in the IEEE 338 standard for measurements of this type and therefore meets the applicable regulatory requirements for performance of response time testing for this set of instrument channels.

3.3 Conclusion Based on the above, the NRC staff finds that the proposed alternative response time testing methodology, which combines observations of the snap-acting limit switch performance during calibration surveillances and the modified response time testing procedure for the components downstream of the MSIV snap-acting limit switch, provides reasonable assurance of detection of response time changes and snap-acting limit switch degradation and, therefore, adequate protection of the public health and safety. The staff also finds that these changes would continue to provide the instrumentation to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, with appropriate controls to maintain these variables and systems within prescribed operating ranges. Therefore, the staff finds that with these changes, the licensee will continue to meet the requirements of 10 CFR Part 50, Appendix A, GDCs 13 and 21 and 10 CFR 50.55a(h). Thus the proposed changes are acceptable.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The NRCs regulations in 10 CFR 50.92 state that the NRC may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91(a), an evaluation of the issue of no significant hazards consideration is presented below.

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

Response: No The proposed amendment would allow for a testing methodology that meets the definition of RPS Response Time. RPS Response Time testing for this function would still be performed to ensure that the full system response would occur to mitigate the consequences of inadvertent MSIV closure as described in USAR section 15.2.4, MSLIV Closures. The proposed amendment does not involve a significant increase in the probability of an accident previously evaluated because the amendment involves no change to the plant or its modes of operation.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed amendment does not change the design, configuration, or method of operation of the plant. The proposed amendment does not involve a physical alteration of the plant (i.e., no new or different kind of equipment will be installed). This request allows for a verification methodology that meets the definition of RPS Response Time.

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

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

Response: No There is no change in safety margin because Technical Specification-required RPS Response Time testing requirements, including total response time, remain unchanged. The proposed amendment does not affect the inputs or assumptions of any of the analyses that demonstrate the integrity of the fuel cladding, reactor coolant system, or containment during accident conditions.

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

Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendment on August 13, 2025. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on August 6, 2025 (90 FR 37894), 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 Commissions 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 Contributors: Youssef Rokes David Rahn Date of Issuance: September 8, 2025

ML25239B557 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EICB/BC NAME BWetzel SLent FSacko DATE 08/27/2025 8/28/2025 08/26/2025 OFFICE OGC/NLO NRR/DORL/LPL3/BC (A)

NRR/DORL/LPL3/PM NAME JWachutka IBerrios BWetzel DATE 09/02/2025 09/08/2025 09/08/2025