ML25076A019

Issuance of Amendment No. 215 Revise Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements

ML25076A019
Person / Time
Site:	Monticello
Issue date:	04/10/2025
From:	Wetzel B NRC/NRR/DORL/LPL3
To:	Hafen S Northern States Power Company, Minnesota
Wetzel B, NRR/DORL/LPL3
References
EPID L-2024-LLA-0135
Download: ML25076A019 (1)

Text

April 10, 2025 Shawn Hafen Site Vice President Northern States Power Company

- Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT NO. 215 RE: REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2024-LLA-0135)

Dear Shawn Hafen:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 215 to Subsequent Renewed Facility Operating License No. DPR-22, for the Monticello Nuclear Generating Plant. The amendment consists of changes to the technical specifications (TSs) in response to your application dated September 24, 2024.

The amendment revises 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 the 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. 50-263

Enclosure:

1. Amendment No. 215 to DPR-22

2. Safety Evaluation cc: Listserv

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 215 Subsequent Renewed License No. DPR-22

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 26, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of 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 Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: April 10, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.04.10 08:05:48 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 215 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-263 Subsequent Renewed Facility Operating License No. DPR-22 Replace the following page of the Subsequent Renewed Facility Operating License No. DPR-22 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

INSERT REMOVE Page 3 Page 3 Technical Specifications Replace the following pages of 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.

INSERT REMOVE 1.1-4 1.1-4 3.4.4-1 3.4.4-1 3.4.4-2 3.4.4-2 Amendment No. 215

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 calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

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

1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, are hereby incorporated in the subsequent renewed license.

NSPM shall operate the facility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall 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)(2). The combined set of plans which contain Safeguards Information protected under 10 CFR 73.21, are entitled: Monticello Nuclear Generating Plant Physical Security, Training and Qualification, and Safeguards Contingency Plan, with revisions submitted through May 12, 2006.

NSPM shall fully implement and maintain in effect all provisions of the Commission-approved Northern States Power Company - Minnesota (NSPM)

Definitions 1.1 Monticello 1.1-4 Amendment No. 215 1.1 Definitions LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell, such as that from pump seals or valve packing that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL TEST logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR) that RATIO (MCPR) exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

RCS Operational LEAKAGE 3.4.4 Monticello 3.4.4-1 Amendment No. 215 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

5 gpm unidentified LEAKAGE;

c.

25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

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. Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

B.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 />

RCS Operational LEAKAGE 3.4.4 Monticello 3.4.4-2 Amendment No. 215 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE increase not within limit.

C.1 Reduce LEAKAGE to within limits.

OR C.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours D. Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increases are within limits.

In accordance with the Surveillance Frequency Control Program

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated September 26, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24270A102), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy, requested changes to the technical specifications (TSs) for Monticello Nuclear Generating Plant (MNGP). 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 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).

1.1 Reactor Coolant System Description Components that contain or transport the coolant to or from the reactor core make up the RCS.

Materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and through operational wear or mechanical deterioration during normal and upset operating environments. Such material degradation could lead to leakage of reactor coolant into containment buildings.

RCS leakage falls under two main categories - identified leakage and unidentified leakage.

Identifying the sources of leakage is necessary for prompt identification of potentially adverse conditions, assessment of safety significance of the leakage, and quick corrective action. A limited amount of leakage from the reactor coolant pressure boundary (RCPB) directly into the containment/drywell atmosphere is expected as the RCS and other connected systems cannot be made 100 percent leak tight. This leakage is detected, located, and isolated from the containment atmosphere so as to not interfere with measurement of unexpected RCS leakage detection.

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. Separation of identified leakage from unidentified leakage provides quantitative information to the operators, allowing them to take corrective action should leakage occur that is detrimental to the safety of the unit and the public.

1.2 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 the term LEAKAGE.

Specifically, the licensee proposed the following changes to adopt TSTF-554:

The TS 1.1 identified LEAKAGE definition a.2 would be revised to remove the exclusion of pressure boundary leakage from identified leakage by deleting either and the phrase not to be pressure boundary LEAKAGE.

The TS 1.1 pressure boundary LEAKAGE definition d 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 STS [Standard Technical Specifications] Bases and added to the definition.

The ACTIONS section of STS 3.4.4, RCS Operational LEAKAGE, would be revised to add a new Condition A to isolate the pressure boundary leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Existing Condition D would be revised to be applicable should any Action of LCO [Limiting Conditions for Operation] 3.4.4 not be met by deleting of Condition A or B.

Existing Conditions A, B, and C would be renumbered to reflect the new Condition A. The existing Condition C would be revised to delete to the condition for when pressure boundary leakage exists because pressure boundary leakage would be addressed by the new Condition A. The Required Actions Associated with existing Conditions A and B would be renumbered accordingly.

1.3 Additional Proposed TS Changes In addition to the changes proposed consistent with the traveler discussed in section 1.1, the licensee proposed the following variations.

Editorial Variations The licensee noted that punctuation in the MNGP TS 1.1 Identified LEAKAGE and Unidentified LEAKAGE definitions did not require changing to align with TSTF-554.

Other Variations The licensee noted that MNGP was not licensed to the 10 CFR 50, appendix A, General Design Criteria (GDC) due to the GDC not being finalized at time of initial licensing of MNGP. MNGP was licensed to the applicable Atomic Energy Commission (AEC) preliminary general design criteria, published July 11, 1967.

2.0 REGULATORY EVALUATION

Per Title 10 of the Code of Federal Regulations (10 CFR) part 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR part 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 CFR part 50.40(a), and those specifically for issuance of operating licenses in 10 CFR part 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.

The regulation at 10 CFR 50.36(c)(2) requires that TSs include 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 RCPB 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 (ML073200271), section B, discussion Leakage Separation, provides information related to separation between identified and unidentified leakage.

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 NRC Standard Technical Specifications, General Electric BWR [Boiling-Water Reactor]/4 Plants NUREG 1433, Volume 1, Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ML212724A357 and ML212724A358, respectively), as modified by NRC-approved travelers.

NUREG-1433 is based on the BWR/4 plant design and is representative of the BWR/3 design.

Traveler TSTF-554 revised the STSs related to RCS operational leakage and the definition of the term LEAKAGE. The NRC approved TSTF-554, under the CLIIP on December 18, 2020 (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 the SRP, chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-554 are applicable to MNGP TSs because the MNGP is a BWR/3, and the NRC staff approved the TSTF-554 changes for BWR/3 designs. The NRC finds that the licensees proposed changes to the MNGP TSs in section 1.1 of this SE are consistent with those found acceptable in TSTF-554.

In the SE of TSTF-554, the NRC staff concluded that TSTF-554 changes to STS 1.1 definition of LEAKAGE and to STS 3.4.4, the LCO addressing conditions and required actions when reactor coolant system pressure boundary leakage exists, are acceptable. The NRC staff found that removing the term nonisolable provides a clearer definition of pressure boundary leakage and that the source of the leakage is 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 identified 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 boundary pressure 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 proposed changes to the TS 1.1 definition clarify what constitutes pressure boundary leakage and the source of leakage does not matter if the TSs have separate limits on pressure boundary leakage and LCO 3.4.4 correctly specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. Also, the NRC staff finds that proposed changes to the Actions of LCO 3.4.4 are adequate remedial actions to be taken until each LCO can be met provide protection to the health and safety of the public. Thus, the proposed changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i) as discussed in section 3.0 of the NRC staffs SE of TSTF-554 3.2 Additional Proposed TS Changes 3.2.1 Editorial The licensee noted that punctuation in the MNGP TS 1.1 Identified LEAKAGE and Unidentified LEAKAGE definitions did not require changing to align with TSTF-554. The NRC staff finds these variations are editorial and do not alter any TS requirements.

3.2.2 Other Variations The licensee noted that MNGP was not licensed to the 10 CFR 50, appendix A, General Design Criteria (GDC) due to the GDC not being finalized at time of initial licensing of MNGP. MNGP was licensed to the applicable AEC preliminary general design criteria, published July 11, 1967.

The NRC staff finds the implementation of TSTF-554 at MNGP will not impact the plants compliance with its licensed design requirements 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 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 amendment on January 24, 2025. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes 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 amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (89 FR 104571). 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.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor(s): C. Rojas, NRR M. Hamm, NRR Date of Issuance: April 10, 2025

ML25076A019 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NAME BWetzel SRohrer SMehta DATE 3/17/2025 3/18/25 2/11/2025 OFFICE OGC/NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME KBernstein IBerrios BWetzel DATE 4/3/2025 4/10/2025 4/10/2025