ML26077A001

Issuance of Amendment No. 203, Technical Specification Change to Allow a One-Time Increase in Main Steam Isolation Valve Allowable Leakage Rate

ML26077A001
Person / Time
Site:	Nine Mile Point  (NPF-069)
Issue date:	03/20/2026
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Mudrick C Constellation Energy Generation
References
EPID L-2026-LLA-0050
Download: ML26077A001 (0)

Text

March 20, 2026 Mr. Christopher H. Mudrick, Sr.

Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 200 Exelon Way Kennett Square, PA 19348

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT 2 - ISSUANCE OF AMENDMENT NO. 203, TECHNICAL SPECIFICATION CHANGE TO ALLOW A ONE-TIME INCREASE IN MAIN STEAM ISOLATION VALVE ALLOWABLE LEAKAGE RATE (EPID L-2026-LLA-0050) (EMERGENCY CIRCUMSTANCES)

Dear Mr. Mudrick:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 203 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit 2 (Nine Mile Point, Unit 2). The amendment consists of changes to the technical specifications (TSs) in response to your application dated March 16, 2026, as supplemented by letter dated March 18, 2026.

Specifically, the amendment revises the Surveillance Requirement (SR) associated with TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), to permit operation of Nine Mile Point, Unit 2, for one operating cycle with: (1) a maximum total main steam line leakage rate limit that is slightly lower than the value assumed in the analysis of record, and (2) a maximum allowable leakage rate limit for one main steam isolation valve that is slightly higher than the value assumed in the analysis of record. This one-time change is applicable only during Nine Mile Point, Unit 2, operating Cycle 21.

The license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations due to the time-critical nature of the amendment.

A copy of the related safety evaluation is enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the NRC staffs final no significant hazards determination. A Notice of Issuance addressing the final no significant

hazards determination and opportunity for a hearing associated with the emergency circumstances will be included in a future biweekly Federal Register notice.

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-410

Enclosures:

1. Amendment No. 203 to NPF-69

2. Safety Evaluation cc: Listserv NINE MILE POINT NUCLEAR STATION, LLC LONG ISLAND LIGHTING COMPANY CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 203 Renewed License No. NPF-69

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated March 16, 2026, as supplemented by letter dated March 18, 2026, 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-69 is hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 203, are hereby incorporated into this license. Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 5 days.

FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Acting 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: March 20, 2026 UNDINE SHOOP Digitally signed by UNDINE SHOOP Date: 2026.03.20 11:57:08 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 203 NINE MILE POINT NUCLEAR STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 4

4 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3.6.1.3-13 3.6.1.3-13 Renewed License No. NPF-69 Amendment 117 through 202 203 (1)

Maximum Power Level Constellation Energy Generation, LLC is authorized to operate the facility at reactor core power levels not in excess of 3988 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 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 203 are hereby incorporated into this license. Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Fuel Storage and Handling (Section 9.1, SSER 4)*

a. Fuel assemblies, when stored in their shipping containers, shall be stacked no more than three containers high.

b. When not in the reactor vessel, no more than three fuel assemblies shall be allowed outside of their shipping containers or storage racks in the New Fuel Vault or Spent Fuel Storage Facility.

c. The above three fuel assemblies shall maintain a minimum edge-to-edge spacing of twelve (12) inches from the shipping container array and approved storage rack locations.

d. The New Fuel Storage Vault shall have no more than ten fresh fuel assemblies uncovered at any one time.

(4)

Turbine System Maintenance Program (Section 3.5.1.3.10, SER)

The operating licensee shall submit for NRC approval by October 31, 1989, a turbine system maintenance program based on the manufacturers calculations of missile generation probabilities.

(Submitted by NMPC letter dated October 30, 1989 from C.D. Terry and approved by NRC letter dated March 15, 1990 from Robert Martin to Mr.

Lawrence Burkhardt, III).

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

PCIVs 3.6.1.3 NMP2 3.6.1.3-13 Amendment 91, 182, 201, SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.12 Verify leakage rate through each MSIV is 39 scfh when tested at 25 psig.*

In accordance with 10 CFR 50 Appendix J Testing Program Plan SR 3.6.1.3.13 Verify combined leakage rate through hydrostatically tested lines that penetrate the primary containment is within limits.

In accordance with 10 CFR 50 Appendix J Testing Program Plan

• During Cycle 21, leakage rate through one MSIV may exceed 39 scfh provided the leakage rate is < 83 scfh and the leakage rate through all four main steam lines is < 155 scfh.

203 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 203 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69 CONSTELLATION ENERGY GENERATION, LLC NINE MILE POINT NUCLEAR STATION, UNIT 2 DOCKET NO. 50-410

1.0 INTRODUCTION

By application dated March 16, 2026 (Agencywide Documents Access and Management System Accession No. ML26075A037), as supplemented by letter dated March 18, 2026 (ML26077A153), Constellation Energy Generation, LLC (CEG, the licensee), requested changes to the technical specifications (TSs) for Nine Mile Point Nuclear Station (Nine Mile Point), Unit 2.

The proposed amendment would revise the surveillance requirement (SR) associated with the Nine Mile Point, Unit 2, TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs).

Specifically, a note would be added to TS SR 3.6.1.3.12, to allow a one-time increase in the maximum allowable leakage rate limit for one main steam isolation valve. The proposed note would also establish a maximum total allowable leakage rate limit for all four main steam lines combined. The one-time update to TS SR 3.6.1.3.12 would apply only during operating cycle 21 (Cycle 21), which is scheduled to end in March 2028. The licensee requested approval of the proposed change as an emergency license amendment in accordance with the regulations at Title 10 of the Code of Federal Regulations (10 CFR) 50.91(a)(5).

2.0 REGULATORY EVALUATION

2.1

System Description

In its letter dated March 16, 2026, Section 2.0, Detailed Description, the licensee described the MSIVs and the need for the requested action. The licensee stated:

The four main steam lines (MSLs), which penetrate the drywell, are automatically isolated by the Main Steam Isolation Valves (MSIVs). There are two MSIVs on each main steam line, one inside containment (i.e., inboard) and one outside containment (i.e., outboard). The MSIVs are functionally part of the primary containment boundary and leakage through these valves provides a potential leakage path for fission products to bypass secondary containment and enter the environment as a ground level release.

The licensee stated that Nine Mile Point, Unit 2, is currently in a refueling outage. During the outage, MSIV local leakage rate testing (LLRT), initially performed on March 12, 2026, identified that two inboard MSIVs exceeded the as-found TS maximum allowable leakage limit for a single valve. The licensee further stated that leakage for one of the affected MSIVs was reduced below the TS limit following maintenance and troubleshooting activities. The requested amendment would allow deferral of repairs for the remaining MSIV (2MSS*AOV6D) until the next scheduled refueling outage in 2028.

The licensee requested that the proposed TS changes be processed on an emergency basis pursuant to 10 CFR 50.91(a)(5). The licensee stated that deferring repair of valve 2MSS*AOV6D until the next refueling outage would reduce radiological dose to plant personnel and avoid a significant delay in plant startup following the current refueling outage.

2.2 Licensees Proposed Changes The licensee proposed to revise SR 3.6.1.3.12 to include a note that provides a temporary change to the SR. The proposed note would allow operation of Nine Mile Point, Unit 2, for one operating cycle (Cycle 21) with a maximum leakage rate of 83 standard cubic feet per hour (scfh) for a single MSIV, when tested at 25 psig. In addition, the proposed note specifies that the total leakage rate for all four main steam lines will not exceed 155 scfh. This change is applicable only during Cycle 21.

SR 3.6.1.3.12 currently states:

Verify leakage rate through each MSIV is 39 scfh when tested at 25 psig.

The TSs specify this SR is to be performed at a frequency in accordance with the 10 CFR Part 50, Appendix J Testing Program Plan.

The licensee proposed the following note to SR 3.6.1.3.12:

• During Cycle 21, leakage rate through one MSIV may exceed 39 scfh provided the leakage rate is 83 scfh and the leakage rate through all four main steam lines is 155 scfh.

2.3 Regulatory Requirements and Guidance The U.S. Nuclear Regulatory Commission (NRC or the Commission) staff reviewed the licensees application to determine whether (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 the activities proposed 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 the health and safety of the public. The NRC staff considered the following regulatory requirements, guidance, and licensing and design-basis information during its review of the proposed changes.

The regulations at 10 CFR 50.36 list the requirements for TSs. Paragraph 50.36(a)(1) states, in part, that each applicant for an operating license shall include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36, Technical specifications.

Paragraph 50.36(b) requires that each license authorizing reactor operation include TSs derived from the analyses and evaluation included in the safety analysis report and amendments thereto. Paragraph 50.36(c) requires that the TSs include items in the following categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements; (4) design features; and (5) administrative controls. Paragraph 50.36(c)(3) states, Surveillance requirements 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.

Section 50.67, Accident Source Term, of 10 CFR Part 50 provides a mechanism for licensed power reactors to replace the traditional source term used in the design-basis accident (DBA) radiological consequence analyses with an acceptable alternative source term (AST). Licensees using the AST are evaluated against the dose criteria specified in 10 CFR 50.67(b)(2):

10 CFR 50.67(b)(2)(i) requires that an individual located at any point on the boundary of the exclusion area (EAB) for any 2-hour period following the onset of the postulated fission product release would not receive a radiation dose in excess of 25 roentgen equivalent man (rem) total effective dose equivalent (TEDE).

10 CFR 50.67(b)(2)(ii) requires that an individual located at any point on the outer boundary of the low population zone (LPZ) who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE.

10 CFR 50.67(b)(2)(iii) requires that adequate radiation protection be provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

In addition, Appendix A, General Design Criteria for Nuclear Power Plants (GDC) to 10 CFR Part 50, Criterion 19, Control room, requires, in part, that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents (LOCAs). Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body for the duration of the accident. GDC 19 further provides that holders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 5 rem TEDE for the duration of the accident.

Section 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, of 10 CFR Part 50 identifies requirements for establishing a program for qualifying electric equipment that is important to safety as defined in 10 CFR 50.49(b). Section 50.49(e)(1) requires that the time-dependent temperature and pressure at the location of the electric equipment important to safety must be established for the most severe DBA during or following which this equipment is required to remain functional. Section 50.49(b)(2) of 10 CFR Part 50 requires qualification of nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs 50.49(b)(1)(A)-(C) by safety-related equipment.

Regulatory Guide (RG) 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ML003716792), provides guidance for meeting requirements in 10 CFR 50.67.

The NRC staffs guidance for review of ASTs is in Standard Review Plan (SRP) Section 15.0.1, Revision 0, Radiological Consequence Analyses Using Alternative Source Terms, dated July 2000 (ML003734190).

The NRC staffs guidance for review of environmental qualification of equipment is in SRP Section 3.11, Revision 3, Environmental Qualification of Mechanical and Electrical Equipment, of NUREG-0800 (ML063600397).

The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, dated March 2010 (ML100351425).

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed change against the requirements and guidance discussed in Section 2.3 of this safety evaluation.

3.1 Radiological Consequences from Increased MSIV Leakage In the current radiological consequence analysis of record for Nine Mile Point, Unit 2, for the Loss of Coolant Design Basis Accident (LOCA DBA), the MSIVs are postulated to leak at a total design leak rate, through two of the four main steam lines with a maximum combined leak rate of 200 scfh, measured at 40 pounds per square inch gauge (psig). In a recent LAR, the licensee requested the leak test pressure to be lowered to 25 psig (ML25029A181), but the analysis of record for the LOCA DBA remains based on the 40 psig test pressure at the time of this application. The DBA LOCA radiological analysis is described in the Nine Mile Point, Unit 2, Updated Safety Analysis Report, Section 15.6.5 (ML24291A165).

The proposed one-time change to the TSs MSIV leakage rate limit for one MSIV, intended to be used for a single cycle (Cycle 21), does not propose to make permanent changes to any design input parameters or assumptions in the MSIV leakage model that was evaluated using the Radionuclide Transport and Removal and Dose Estimation (RADTRAD) computer code. The licensees LAR provides the analysis of record post-LOCA doses at the three receptor locations, Control Room (CR), Exclusion Area Boundary (EAB), and the Low Population Zone (LPZ), and provides an update to the post-LOCA dose analysis. The current calculated post-LOCA doses provided in this LAR using the RADTRAD code are 2.25 rem TEDE (control room), 1.06 rem TEDE (EAB), and 0.91 rem TEDE (LPZ).

The NRC staff performed independent sensitivity studies utilizing RADTRAD to evaluate the dose consequence from the proposed increase in leakage associated with valve 2MSS*AOV6D and hypothetical increases in leakage beyond what is requested in the LAR. These sensitivity studies address the margin associated with the dose analysis.

The NRC staffs sensitivity studies retained the post-LOCA dose contributions from all other dose pathways (containment leakage, ESF (Engineered Safety Feature) leakage, etc.) as provided in the LAR, but increased the leak rate from the main steam line associated with 2MSS*AOV6D. The purpose of the sensitivity studies was to provide reasonable assurance that in the event the valve 2MSS*AOV6D exceeded the proposed new TS leak rate by some amount, the dose limits in the CR and at the EAB and LPZ would still remain within regulatory limits. A bounding study was performed where the leakage from the main steam line associated with valve 2MSS*AOV6D doubled from the analysis of record. The staffs analysis demonstrated that, even under conditions which assume leakage significantly greater than the value requested for valve 2MSS*AOV6D, the calculated doses at the three receptor locations remain within applicable regulatory limits with substantial margin. It should be noted that these calculations assume that that the downstream valve 2MSS*AOV7D fails to close and mitigate the accident.

With respect to the defense in depth associated with a finding of reasonable assurance of adequate protection, the NRC staff notes that the valve in series with 2MSS*AOV6D is the outboard valve 2MSS*AOV7D. This valve was tested during the current outage N2R20, and the test yielded an as found leak rate of 1.95 scfm. During the previous outage N2R19 the valve 2MSS*AOV7D leak test yielded a 1.87 scfm leak rate. These results demonstrate that the downstream valve, which provides additional isolation in series with 2MSS*AOV6D, has exhibited minimal leakage and only nominal degradation in leak tightness over the previous operating cycle. The NRC staffs sensitivity studies, described above, did not assume this valve was operable.

Radiological Consequences Conclusion The NRC staff reviewed the proposed TSs change as it relates to the DBA LOCA radiological analysis and determined that it meets the dose criteria in 10 CFR 50.67(b)(2) and the requirements of GDC 19 in Appendix A to 10 CFR Part 50, because the proposed increase in TS leak rate for one line and the related accident dose consequence analysis was found to be within the regulatory limits for the control room, exclusion area boundary, and the low population zone. Additionally, the change to the TSs as related to the MSIV leakage model was reviewed for consistency with the guidance in RG 1.183, Revision 0, and SRP Section 15.0.1.

Considering the information provided in the LAR and its supplement, the sensitivity studies performed by NRC staff, and the staffs evaluation of the defense in depth associated with valve 2MSS*AOV6D, the NRC staff determined there is reasonable assurance that (1) GDC 19 to Appendix A to 10 CFR Part 50 will continue to be met, and (2) the dose criteria in 10 CFR 50.67(b)(2) will continue to be met. Therefore, the proposed revision, limited to the next operating Cycle 21, to TS 3.6.1.3.12 is acceptable.

3.2 Environmental Qualification The NRC staff evaluated whether equipment and components would remain bounded by the existing environmental qualification (EQ) as a result of the proposed change.

In the LAR, the licensee stated that the EQ program has been evaluated for both chemical-mechanical and radiological impacts from MSIV leakage. According to the licensee, the equipment and components potentially impacted by the modified MSL leak rate are located in the main steam tunnel and turbine building, downstream of the MSLs. The bounding accident temperature and pressure profiles in the main steam tunnel and turbine building are associated with the high energy line break (HELB). In addition, the scenario evaluated for the HELB has significantly more moisture and heat impact on the equipment than the MSIV leakage would impose. Furthermore, the licensee stated that the zone radiation calculations already incorporate a combined leakage rate of 200 scfh. Therefore, the proposed increase in allowable MSIV leakage would contribute no additional environmental impact to equipment qualified for use in the main steam tunnel or the turbine building. Based on its review of the LAR, the NRC staff finds that the licensee has demonstrated that temperature, pressure, humidity, and radiation conditions for the requested amendment remain bounded by the existing EQ for electrical equipment.

During its review, the NRC staff noted that it was unclear whether the licensee had evaluated the impact of the proposed change on EQ margins required by 10 CFR 50.49(e)(8). The NRC staff requested that the licensee confirm that these EQ margins would not be adversely affected. In its letter dated March 18, 2026, the licensee stated that the impact of the proposed change to the required environmental qualification margins was considered and, based on the proposed changes in maximum MSL leakage, the margins associated with the requirements of 10 CFR 50.49(e)(8) are not impacted. Based on this information, the NRC staff finds that the licensee has provided reasonable assurance that the proposed change will not adversely affect EQ margins required by 10 CFR 50.49(e)(8).

The licensee also stated that since there is no change to EQ design-basis temperatures, pressure, humidity, or radiation values, the proposed change in MSIV leakage has no impact on nonsafety-related equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions by the safety-related equipment. In addition, the licensee stated that there are no components that are being added to the EQ equipment list due to the proposed change in allowable MSIV leakage. Based on its review of the licensees response, the staff finds that the licensee has provided reasonable assurance that the proposed change will not adversely affect the potential for nonsafety-related electrical equipment to prevent satisfactory accomplishment of safety functions, and that no new electrical equipment needs to be added to the licensees 10 CFR 50.49 EQ program as a result of the proposed change to the MSIV leakage rate, because the proposed change will not result in an impact to equipment beyond what has previously been analyzed.

Environmental Qualification Conclusion Based on its review of the information in the LAR, as supplemented, the NRC staff concludes that the proposed changes will not adversely affect the Nine Mile Point, Unit 2, EQ program and will continue to meet the requirements of 10 CFR 50.49.

3.3 Changes to TS 3.6.1.3, Primary Containment Isolation Valves, SR 3.6.1.3.12 The licensee proposed a revision to SR 3.6.1.3.12 to include a note that provides a temporary change to TS SR 3.6.1.3.12. Specifically, the proposed note would allow operation of Nine Mile Point, Unit 2, for one operating (Cycle 21) with a maximum single MSIV leakage of 83 scfh when tested at 25 psig. The note also applies a requirement that the total leakage rate for all main steam lines will not exceed 155 scfh. The note is applicable for a single cycle and will expire upon shutdown for the next scheduled refueling outage.

The NRC staff reviewed the information in the licensees LAR and performed independent confirmatory calculations. Based on its assessment, the NRC staff concluded that that proposed change will continue to ensure safe operation of the facility. The NRC staff finds the proposed changes to TS SR 3.6.1.3.12 with a note allowing the temporary change described above, to be acceptable. The temporary change to the SR provides reasonable assurance that the necessary quality of systems and components is maintained, that facility operation remains within safety limits, and that the LCO is met, consistent with the requirements of 10 CFR 50.36(c)(3). Based on its review as discussed above, the NRC staff has determined that TS SR 3.6.1.3.12, as amended by the proposed one-time note, will continue to meet the requirements of 10 CFR 50.36(b) because it is based on the analyses and evaluations included in the safety analysis report and amendments thereto and complies with applicable regulatory requirements.

Technical Conclusion Based on the evaluation discussed above, the NRC staff concludes that TS 3.6.1.3, as amended by the proposed change, continues to meet the requirements of 10 CFR 50.36 and will ensure safe operation of the facility within applicable radiological dose limits. Accordingly, the NRC staff finds the proposed amendment acceptable.

Additionally, the NRC staff reviewed the proposed TS changes for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff determined that the proposed changes are acceptable, as they are consistent with the existing terminology and format of the Nine Mile Point, Unit 2, TSs.

4.0 EMERGENCY SITUATION

=

Background===

The NRCs regulations in 10 CFR 50.91(a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, the NRC may issue a license amendment that involves no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing an opportunity for hearing and for public comment after issuance.

As discussed in the licensees application dated March 16, 2026, as supplemented on March 18, 2026, the licensee requested that the proposed amendment be processed by the NRC on an emergency basis. The licensee provided the following basis to support the finding that an emergency situation exists:

NMP2 is currently in a refueling outage. During the outage MSIV LLRTs first performed on March 12, 2026, two of the inboard MSIVs each exceeded their as-found current TS maximum allowable limit for a single valve. One MSIV, 2MSS*AOV6A (located inside primary containment) failed the LLRT with a measured leakage rate of approximately 195 scfh. Leakage through the 2MSS*AOV6A MSIV was lowered below the current TS maximum allowable limit for a single valve through the performance of extensive maintenance and troubleshooting. A second MSIV, 2MSS*AOV6D (located inside primary containment) failed the LLRT with a measured leakage of approximately 81 scfh.

Additionally, recirculation loop discharge valve 2RCS*MOV18B failed to close on March 9, 2026. The requested TS amendment supports deferral of repairs on 2MSS*AOV6D until the next NMP2 refueling outage in 2028. The proposed changes are being requested on an emergency basis pursuant to 10 CFR 50.91(a)(5).

Based on the current refueling outage plan, this license amendment request would need to be implemented on or before March 23, 2026, in order to prevent affecting outage critical path and causing the delay of NMP2 startup. Failure of multiple MSIVs and an additional large valve (2RCS*MOV18B) was not foreseen based on historical LLRT results for MSIVs performed in previous refueling outages. In-body MSIV maintenance, which would be the next required maintenance technique required to lower 2MSS*AOV6D leakage, also requires specialized vendor support. Lowering 2MSS*AOV6D leakage through in-body valve maintenance would delay startup by approximately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> and incur an additional approximately 2.15 person-rem of radiological dose. Postponing repair of 2MSS*AOV6D until the next refueling outage would reduce plant staff radiological dose accumulation and prevent a significant delay to startup following the current refueling outage.

Further, NRC approval is needed sooner than can be provided under exigent circumstances, and this license amendment request is timely considering the unplanned nature of the issues and time required to develop the technical justification supporting the proposed change. Therefore, CEG has determined that emergency circumstances exist and that NMP2 did not knowingly cause the emergent situation. Further, CEG affirms a best effort has been made for a timely license amendment application. Accordingly, CEG requests an expedited review of the proposed emergency license amendment in accordance with the provisions of 10 CFR 50.91(a)(5) to avoid the unnecessary delay of NMP2 startup following the current refueling outage.

NRC Staff Conclusion The NRC staff has reviewed the licensees justification for processing the proposed amendment as an emergency amendment, as discussed above. Nine Mile Point, Unit 2, is currently in a planned outage. If the amendment is not approved by the requested date of March 22, 2026, the unit would be unable to return to power operation because the measured MSIV leakage would exceed the TS allowed limit in the modes of operation needed to return to power (Modes 1, 2, and 3).

Accordingly, the NRC staff finds that an emergency situation exists consistent with the provisions of 10 CFR 50.91(a)(5). The NRC staff further finds that: (1) the licensee used its best efforts to make a timely application, (2) the licensee could not reasonably have avoided the situation, and (3) the licensee has not abused the provisions of 10 CFR 50.91(a)(5).

Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff concludes that a valid need exists to issue the license amendment under the emergency provisions of 10 CFR 50.91(a)(5).

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, 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.

The licensees evaluation of no significant hazards consideration is presented below:

1.

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

Response: No.

The changes in Main Steam Isolation Valve (MSIV) leakage rate limits for a single cycle have been assessed against the radiological consequence analysis of the Loss of Coolant Accident (LOCA). Based on the results of the assessment, it has been demonstrated that, with the requested changes, the dose consequences of the currently approved LOCA analysis remain bounding and are within the acceptance criteria provided by the NRC for use with the Alternative Source Term (AST) methodology in 10 CFR 50.67.

The proposed changes to MSIV leakage limits do not involve a physical change to any plant structure, system, or component. As a result, no new failure modes of the MSIVs have been introduced. The changes affect leakage limit assumptions that constitute inputs to the evaluation of the consequences. The radiological consequences of the analyzed LOCA have been evaluated using the plant licensing basis for this accident. The evaluation concludes that the currently approved LOCA analysis bounds the proposed Technical Specification change. Adequate margin to the regulatory limits specified in 10 CFR 50.67 for control room and offsite doses is still available.

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

2.

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

Response: No.

These changes do not affect the design or operation of any component in the facility such that new or different equipment failure modes are created. When a single MSIV in a line exceeds the NMP2 criteria for Local Leak Rate Test (LLRT) leakage results, a condition report is generated, and remediation is controlled by the corrective action program to restore compliance. As such, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously 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.

This proposed license amendment involves changes in the MSIV leakage rate limits. The MSIV leakage rate limits are used in the analysis of the LOCA radiological consequences. The analysis has been performed using conservative methodologies.

Safety margins and analytical conservatisms have been evaluated and have been found to adequately address the effects of the proposed MSIV leakage limits. The analyzed LOCA event has been selected and margin has been retained to ensure that the analysis adequately bounds the postulated event scenario. The dose consequences of this limiting event are within the acceptance criteria presented in 10 CFR 50.67 for control room and offsite doses. The margin of safety is that provided by meeting the applicable regulatory limits.

Therefore, the proposed change 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.

6.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on March 17, 2026. On March 19, 2026, the State official confirmed that the State of New York had no comments.

7.0 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 and changes surveillance requirements. 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 made a finding that the amendment involves no significant hazards consideration. 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.

8.0 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: S. Meighan S. Smith H. Wagage J. Cintron Date: March 20, 2026

PKG: ML26077A207 Document: ML26077A001 e-Concurrence case: 20260318-1014