ML26015A145
| ML26015A145 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/27/2026 |
| From: | Wetzel B NRC/NRR/DORL/LPL3 |
| To: | Paulhardt W Northern States Power Company, Minnesota |
| Wetzel B, NRR/DORL/LPL3 | |
| References | |
| EPID L-2025-LLA-0142 | |
| Download: ML26015A145 (0) | |
Text
January 27, 2026 Mr. Werner Paulhardt Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -
ISSUANCE OF AMENDMENTS NOS. 249 AND 237 RE: REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2025-LLA-0142)
Dear Mr. Paulhardt:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 249 to Renewed Facility Operating License No. DPR-42 and Amendment No. 237 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated September 3, 2025.
The amendments revise the TSs related to reactor coolant system operational leakage and the TS definition of LEAKAGE, based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements.
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/
Beth Wetzel, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosures:
- 1. Amendment No. 249 to DPR-42
- 2. Amendment No. 237 to DPR-60
- 3. Safety Evaluation cc: Listserv
NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 249 Renewed License No. DPR-42
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (NSPM, the licensee), dated September 3, 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, 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. DPR-42 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 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
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 27, 2026 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2026.01.27 12:09:49 -05'00'
NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 Renewed License No. DPR-60
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (NSPM, the licensee), dated September 3, 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, 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. DPR-60 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 237, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 27, 2026 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2026.01.27 12:08:54 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 249 AND 237 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.
Renewed Facility Operating License No. DPR-42 REMOVE INSERT Page 3 Page 3 Renewed Facility Operating License No. DPR-60 REMOVE INSERT Page 3 Page 3 Technical Specifications 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 1.1-3 1.1-3 1.1-4 1.1-4 1.1-5 1.1-5 1.1-6 1.1-6 1.1-7 1.1-7 1.1-8 1.1-8 3.4.14-1 3.4.14-1 3.4.14-2 3.4.14-2 Renewed Operating License No. DPR-42 Amendment No. 249 (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:
Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, are hereby incorporated in the renewed operating license.
NSPM shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 237 (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purposes of volume reduction and decontamination.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:
Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 237, are hereby incorporated in the renewed operating license.
NSPM shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-3 Unit 2 - Amendment No. 237 1.1 Definitions (continued)
DOSE DOSE EQUIVALENT XE-133 shall be that concentration of EQUIVALENT Xe-133 (microcuries per gram) that alone would produce the XE-133 same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."
LEAKAGE LEAKAGE from the Reactor Coolant System (RCS) shall be:
- a.
Identified LEAKAGE
- 1.
LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or
- 3.
RCS LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
- b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-4 Unit 2 - Amendment No. 237 1.1 Definitions (continued)
LEAKAGE
- c.
Pressure Boundary LEAKAGE (continued)
LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.
MASTER A MASTER RELAY TEST shall consist of energizing all RELAY master relays in the channel required for channel OPERABILITY TEST and verifying the OPERABILITY of each required master relay.
The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLE-A system, subsystem, train, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-5 Unit 2 - Amendment No. 237 1.1 Definitions (continued)
PHYSICS PHYSICS TESTS shall be those tests performed to measure the TESTS fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a.
Described in Appendix J of the USAR, Pre-Operational and Startup Tests;
- b.
Authorized under the provisions of 10 CFR 50.59; or
- c.
Otherwise approved by the Nuclear Regulatory Commission.
PRESSURE AND The PTLR is the unit specific document that provides the reactor TEMPERATURE vessel pressure and temperature limits, including heatup and LIMITS cooldown rates, and the OPPS arming temperature for the current REPORT reactor vessel fluence period. These pressure and temperature limits (PTLR) shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these operating limits is addressed in LCO 3.4.3, RCS Pressure and Temperature (P/T)
Limits, LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -Reactor Coolant System Cold Leg Temperature (RCSCLT)
> Safety Injection (SI) Pump Disable Temperature, and LCO 3.4.13, Low Temperature Overpressure Protection (LTOP)
- Reactor Coolant System Cold Leg Temperature (RCSCLT)
< Safety Injection (SI) Pump Disable Temperature.
QUADRANT QPTR shall be the ratio of the maximum upper excore detector POWER TILT calibrated output to the average of the upper excore detector RATIO (QPTR) calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED RTP shall be a total reactor core heat transfer rate to the reactor THERMAL coolant of 1677 MWt.
POWER (RTP)
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-6 Unit 2 - Amendment No. 237 1.1 Definitions (continued)
REACTOR The RTS RESPONSE TIME shall be that time interval from when TRIP the monitored parameter exceeds its RTS trip setpoint at the channel SYSTEM (RTS) sensor output until opening of a reactor trip breaker. The response
RESPONSE
time may be measured by means of any series of sequential, TIME 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 have been evaluated in accordance with an NRC approved methodology.
SHUTDOWN SDM shall be the instantaneous amount of reactivity by which:
MARGIN (SDM)
- a.
The reactor is subcritical; or
- b.
The reactor would be subcritical from its present condition assuming all rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design temperature.
SLAVE A SLAVE RELAY TEST shall consist of energizing all slave relays RELAY in the channel required for channel OPERABILITY and verifying TEST the OPERABILITY of each required slave relay. The SLAVE shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlappping, or total steps.
THERMAL THERMAL POWER shall be the total reactor core heat transfer POWER rate to the reactor coolant.
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-7 Unit 2 - Amendment No. 237 1.1 Definitions (continued)
TRIP A TADOT shall consist of operating the trip actuating device and ACTUATING verifying the OPERABILITY of all devices in the channel DEVICE required for trip actuating device OPERABILITY. The TADOT OPERATIONAL shall include adjustment, as necessary, of the trip actuating device TEST so that it actuates at the required setpoint within the necessary (TADOT) accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.
Definitions 1.1 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 1.1-8 Unit 2 - Amendment No. 237 Table 1.1-1 (page 1 of 1)
MODES MODE TITLE REACTIVITY CONDITION (keff)
% RATED THERMAL POWER(a)
AVERAGE REACTOR COOLANT TEMPERATURE
(°F) 1 Power Operation
> 0.99
> 5 NA 2
Startup
> 0.99
< 5 NA 3
Hot Standby
< 0.99 NA
> 350 4
Hot Shutdown(b)
< 0.99 NA 350 > Tavg > 200 5
Cold Shutdown(b)
< 0.99 NA
< 200 6
Refueling(c)
NA NA NA (a)
Excluding decay heat.
(b)
All reactor vessel head closure bolts fully tensioned.
(c)
One or more reactor vessel head closure bolts less than fully tensioned.
RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 3.4.14-1 Unit 2 - Amendment No. 237 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4.14 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE;
- b. 1 gpm unidentified LEAKAGE;
- c. 10 gpm identified LEAKAGE; and
- d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary LEAKAGE exists.
A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Required Action and associated Completion Time of Condition A not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours
RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. 249 Units 1 and 2 3.4.14-2 Unit 2 - Amendment No. 237 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. RCS unidentified LEAKAGE not within limit.
C.1 Reduce LEAKAGE to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. Required Action and associated Completion Time of Condition C not met.
D.1 Be in MODE 3.
AND D.2.1 Identify LEAKAGE.
OR D.2.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 54 hours 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> E. RCS identified LEAKAGE not within limit for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
E.1 Be in MODE 3.
AND E.2.1 Reduce LEAKAGE to within limits.
OR E.2.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 14 hours 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> F. Primary to secondary LEAKAGE not within limit.
F.1 Be in MODE 3.
AND F.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 249 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By application dated September 3, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25247A026), Northern States Power Company, a Minnesota Corporation (the licensee), requested changes to the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. The proposed changes would revise the TSs related to reactor coolant system (RCS) operational leakage and the definition of the term LEAKAGE based on Technical Specifications Task Force (TSTF)
Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, (TSTF-554)
(ML20016A233), and the associated U.S. Nuclear Regulatory Commission (NRC or the Commission) staff safety evaluation (SE) of TSTF-554 (ML20322A024). In its application, the licensee requested that the NRC process the proposed amendment under the Consolidated Line Item Improvement Process (CLIIP).
Components that contain or transport the coolant to or from the reactor core make up the RCS.
RCS component joints are made by welding, bolting, rolling, or pressure loading. Valves isolate connecting systems from the RCS.
During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration. The purpose of the RCS operational leakage specification is to limit system operation in the presence of leakage to amounts that do not compromise safety. Prairie Island TS 3.4.14 specifies the types and amounts of leakage.
The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Quickly separating the identified leakage from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.
A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100 percent leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.
1.1 Proposed TS Changes to Adopt TSTF-554 In accordance with NRC staff-approved TSTF-554, the licensee proposed changes that would revise the TSs related to RCS operational leakage and the definition of leakage. Specifically, the licensee proposed the following changes to adopt TSTF-554:
The TS 1.1 LEAKAGE definition for Identified LEAKAGE, item a.2, would be revised to remove the exclusion of pressure boundary leakage by deleting either and the phrase or not to be pressure boundary LEAKAGE.
The TS 1.1 LEAKAGE definition for Pressure Boundary LEAKAGE, item c, would be revised to delete the word nonisolable. The sentence, LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE, would be relocated from the Bases and added to the definition.
Additionally, the LEAKAGE definition would be revised by other editorial changes to reflect the deletion and listed definitions.
The Actions section of TS 3.4.14, RCS Operational LEAKAGE, would be revised as follows:
o New Action A would be added to address the Condition when pressure boundary LEAKAGE exists with a Required Action to isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within a Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1.2 Additional Proposed TS Changes In addition to the changes proposed consistent with the Traveler discussed in Section 1.1 of this SE, the licensee proposed several variations. These variations are described in license amendment request (LAR) Section 2.2, Variations. For example, the licensee noted that Prairie Island TS has different numbering than the standard technical specifications (STS) for Westinghouse plants. Specifically, the Prairie Island TS section number for RCS leakage is 3.4.14 versus 3.4.13 used in TSTF-554. There are also minor format changes made for consistency with the STS approach. Furthermore, Prairie Island TS 3.4.14 contains some different Required Actions and Completion Times in comparison to the STS; however, NRC-approved TSTF-554 changes did not materially impact these items, as described in LAR Section 2.2.
In addition to the variations described above, there are other variations that stem from the fact that Prairie Islands current TS 3.4.14 format contains four Conditions that encompass the two Conditions of STS 3.4.13 on which TSTF-554 was based. When Prairie Island elected to pursue adoption of TSTF-554, this difference resulted in adding one additional new Condition to Prairie Island TS 3.4.14. Specifically, the STS as modified by TSTF-554 addressed the pressure boundary leakage aspect of RCS Operational LEAKAGE (i.e., the purpose for TSTF-554) by adding one new Condition and appropriately modifying the existing two Conditions, resulting in a total of three Conditions. In comparison, to address the pressure boundary leakage aspect of RCS Operational LEAKAGE and meet the intent of TSTF-554, the Prairie Island TS would add two new Conditions and appropriately modify the existing four Conditions, ending up with a total of six Conditions. As such, the Actions section of Prairie Island TS 3.4.14 would be revised as follows:
New Action B would be added to address the Condition when the Required Action and associated Completion Time of Condition A not met with a Required Action to be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Existing Actions A through D would be renumbered C through F, respectively, to reflect the new Actions A and B. The existing Condition D would be revised to delete the condition for when pressure boundary leakage exists because pressure boundary leakage would be addressed by the new Actions A and B.
Lastly, one of the items presented in LAR Section 2.2, Variations, identifies that the Traveler, TSTF-554, describes applicable regulatory requirements to include Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants. Prairie Island was not licensed to the 10 CFR 50, Appendix A, GDC.
2.0 REGULATORY EVALUATION
The regulation at 10 CFR 50.36(c)(2) requires that TSs include limiting conditions for operation (LCOs). Per 10 CFR 50.36(c)(2)(i), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
The regulation at 10 CFR 50.2 defines reactor coolant pressure boundary in part as all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves. Regulatory Guide (RG) 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, dated May 2008, Section B, Discussion, on Leakage Separation, provides information related to separation between identified and unidentified leakage (ML073200271).
The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), March 2010 (ML100351425).
As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the NUREG-1431, Revision 51, as modified by NRC-approved travelers. Traveler TSTF-554 revised the STSs 1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1 Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ADAMS Accession Nos.
ML21259A155 and ML21259A159, respectively).
related to RCS operational leakage and the definition of the term LEAKAGE. The NRC approved TSTF-554, under the CLIIP on December 18, 2020 (Package, ML20324A083).
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes to Adopt TSTF-554 The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-554. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-554 are applicable to Prairie Island, Units 1 and 2, TSs because Prairie Island, Units 1 and 2, are Westinghouse pressurized-water reactors (PWRs) and the NRC staff approved the TSTF-554 changes for Westinghouse PWR designs. The NRC finds that the licensees proposed changes to the Prairie Island, Units 1 and 2, TSs in Section 1.1 of this SE are consistent with those found acceptable in TSTF-554.
In the SE for TSTF-554, the NRC staff concluded that TSTF-554 changes to STS 1.1 definition of LEAKAGE and to STS 3.4.13, which addressed conditions and required actions when reactor coolant system pressure boundary leakage exists, were acceptable. The NRC staff found that removing the term nonisolable provided a clearer definition of pressure boundary leakage and that the source of the leakage was not relevant to this capability provided that separate, appropriate limits on pressure boundary leakage have been established. Therefore, the proposed change to the definition of leakage was acceptable as it did not conflict with 10 CFR 50.2 and was consistent with RG 1.45. The NRC staff further found that proposed new Condition A on pressure boundary leakage, including its associated Required Action A.1 and Completion Time, acceptable because the LCO revisions continue to specify the lowest functionable capability of equipment, identify remedial actions and require shutdown of the reactor if the remedial actions cannot be met.
The NRC staff finds that the proposed changes to the Prairie Island, Units 1 and 2 TS are acceptable because the licensees proposed changes are consistent with those found acceptable in TSTF-554. Specifically, the proposed changes to TS 1.1 leakage definition clarifies what constitutes pressure boundary leakage and LCO 3.4.14 correctly specifies the lowest functional capability or performance levels of equipment required for safe operation of the facility. In addition, the NRC staff finds that the proposed changes to the Actions of TS 3.4.14 provide adequate remedial actions to be taken until the LCO can be met and therefore provide protection to the health and safety of the public. Thus, the NRC staff finds the proposed changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i).
3.2 Additional Proposed TS Changes The licensee noted that Prairie Island, Units 1 and 2, proposed TS changes contained several variations. These variations are described in LAR Section 2.2 and are summarized in SE Section 1.2.
The NRC staff finds that the proposed variations, such as different TS numbering and format differences are acceptable because they do not substantively alter TS requirements. In addition, variations such as adding a new Condition B, revising existing Condition D to delete the condition for when pressure boundary leakage exists because pressure boundary leakage is addressed by the new Conditions A and B, and renumbering the existing conditions are acceptable because they alter TS requirements consistent with the intent of TSTF-554 and retain the current Prairie Island TS 3.4.14 content and structure, where appropriate.
LAR Section 1.2 identifies that the Traveler discusses the applicable regulatory requirements and guidance, including 10 CFR Part 50, Appendix A, GDC. Prairie Island was not licensed to the 10 CFR 50, Appendix A, GDC. The Prairie Island Updated Safety Analysis Report (USAR),
Section 1.5, General Design Criteria, (ML24128A105) states in part, the following:
The Prairie Island Nuclear Generating Plant was designed and constructed to comply with NSPs [Northern States Power] understanding of the intent of the proposed AEC [Atomic Energy Commission] General Design Criteria for Nuclear Power Plant Construction Permits (Appendix A to 10CFR50), as published in the Federal Register on July 11, 1967 Since the construction of the plant was significantly completed prior to the issuance of the February 20, 1971, 10CFR50, Appendix A General Design Criteria, the plant was not reanalyzed and the FSAR
[final safety analysis report] was not revised to reflect these later criteria.
However, the AEC Safety Evaluation Report acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and... are satisfied that the plant design generally conforms to the intent of these criteria.
The NRC staff reviewed the USAR discussion provided above and the TSTF-554 Traveler.
The NRC staff finds that TSTF-554 referenced GDCs 14 and 30, which provide requirements for reactor coolant pressure boundary and monitoring reactor coolant leakage, respectively. The NRC staff reviewed the Prairie Island USAR Section 1.5, and finds that it discusses GDCs 9 and 33, which provide requirements for reactor coolant pressure boundary, and GDC 16, which provides requirements for monitoring reactor coolant leakage. The NRC staffs review confirms the USAR statement cited above that the plant design generally conforms to the intent of these criteria. Based on these findings, the NRC staff concludes that that these differences in GDC are not substantive with respect to the proposed change to TS 3.4.14 and do not affect the applicability of TSTF-554 to Prairie Island.
3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes, where appropriate, are consistent with Chapter 16.0 of the SRP and are therefore acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Minnesota State official was notified of the proposed issuance of the amendments on December 29, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, published in the Federal Register on September 30, 2025 (90 FR 46933), and there has been no public comment on such finding.
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: C. Ashley, NRR Date of Issuance: January 27, 2026
ML26015A145 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NAME BWetzel SLent SMehta DATE 01/20/2026 01/20/2026 12/19/2026 OFFICE NRR/DORL/LPL3/BC (A)
NRR/DORL/LPL3/PM NAME IBerrios BWetzel DATE 01/27/2026 01/27/2026