ML26062A215
| ML26062A215 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  (NPF-035, NPF-052, NPF-017, NPF-009) |
| Issue date: | 03/25/2026 |
| From: | Klos L Plant Licensing Branch II |
| To: | Gibby S Duke Energy |
| Stone Z | |
| References | |
| EPID L-2025-LLA-0132, 20260303-50016 | |
| Download: ML26062A215 (0) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION March 25, 2026 Mr. Shawn Gibby Vice President Nuclear Engineering Duke Energy 525 S. Tyron St.
Charlotte, NC 28202
SUBJECT:
CATAWBA NUCLEAR STATION, UNITS 1 AND 2, AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 323, 319; 333, 312, RESPECTIVELY, TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW USAGE OF A FULL SPECTRUM LOSS-OF-COOLANT-ACCIDENT (LOCA)
METHODOLOGY (EPID L-2025-LLA-0132)
Dear Mr. Gibby:
The U.S. Nuclear Regulatory Commission has issued the following enclosed Amendment Nos. 323 and 319 to Renewed Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station (Catawba), Units 1 and 2, and the following enclosed Amendment Nos. 333 and 312 to Renewed Facility Operating License Nos. NPF-9 and NPF17 for the McGuire Nuclear Station (McGuire), Units 1 and 2, respectively. The amendments are in response to your application RA-25-0014 dated August 18, 2025, as supplemented by letter RA-25-0280 dated January 8, 2026.
The amendments revise Catawbas and McGuires Technical Specifications (TSs) to include Westinghouse topical report WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology), to the list of NRC-approved analytical methods used to determine the core operating limits in TS 5.6.5, CORE OPERATING LIMITS REPORT. The amendments also annotate select legacy LOCA methods and the deletion of others to restrict their future use and allow for a staggered implementation during refueling outages at each unit.
The amendments also revise TSs 4.2.1, Fuel Assemblies, to permit the use of the Westinghouse fuel cladding alloy designated as AXIOM. Specifically, the proposed amendment requests a revision to the TSs to update the description of fuel assemblies specified in TS 4.2.1, Fuel Assemblies, and add the Westinghouse topical report WCAP-18546-P-A, Westinghouse AXIOM Cladding for use in Pressurized Water Reactor Fuel to the referenced NRC-approved analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding.
NOTICE: Enclosure 6 to this letter contains Proprietary Information. When separated from Enclosure 6, this document is DECONTROLLED.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION S. Gibby OFFICIAL USE ONLY - PROPRIETARY INFORMATION The adoption of WCAP-16996-P-A and WCAP-18545-P-A in the COLR is only for use by Catawba, Unit 1, and McGuire, Units 1 and 2. The licensee plans to submit a separate request for Catawba, Unit 2, at a later date. The deletion of legacy LOCA analysis methods is administrative in nature and is applicable to Catawba, Units 1 and 2, and McGuire Units 1 and 2.
The NRC staff has determined that the related safety evaluation contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, requests for withholding. The proprietary information is indicated by text enclosed with double brackets. The proprietary version of the safety evaluation is provided as. Accordingly, the NRC staff has also prepared a non-proprietary version of the safety evaluation, which is provided in Enclosure 5.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions Federal Register notice.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION S. Gibby OFFICIAL USE ONLY - PROPRIETARY INFORMATION If you have any questions, please contact me at (301) 415-0610 or via email at John.Klos@nrc.gov.
Sincerely,
/RA/
John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413, 50-414 50-369, 50-370
Enclosures:
- 1. Amendment No. 323 to NPF-35
- 2. Amendment No. 319 to NPF-52
- 3. Amendment No. 333 to NPF-9
- 4. Amendment No. 312 to NPF-17
- 5. Public Non-proprietary Safety Evaluation
- 6. Non-Public Proprietary Safety Evaluation cc: w/o Enclosure 6 Listserv
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 323 Renewed License No. NPF-35
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC (licensee), dated August 18, 2025, as supplemented by letter dated January 8, 2026, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 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-35 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 323, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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 prior to the Unit 1 Cycle 30 reload campaign, currently scheduled for Spring 2026.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-35 and Technical Specifications Date of Issuance: March 25, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.25 15:07:39 -04'00'
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 319 Renewed License No. NPF-52
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC (licensee), dated August 18, 2025, as supplemented by letter dated January 8, 2026, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 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-52 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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 prior to the Unit 1 Cycle 30 reload campaign, currently scheduled for Spring 2026.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-52 and the Technical Specifications Date of Issuance: March 25, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.25 15:08:18 -04'00'
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 333 Renewed License No. NPF-9
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-9 filed by the Duke Energy Carolinas, LLC (licensee), dated August 18, 2025, as supplemented by letter dated January 8, 2026, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 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-9 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 333, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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 prior to the Unit 1 Cycle 32 reload campaign, currently scheduled for Fall 2026.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-9 and Technical Specifications Date of Issuance: March 25, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.25 15:08:55 -04'00'
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 312 Renewed License No. NPF-17
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-17 filed by the Duke Energy Carolinas, LLC (licensee), dated August 18, 2025, as supplemented by letter dated January 8, 2026, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 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-17 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 312, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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 prior to the Unit 1 Cycle 32 reload campaign, currently scheduled for Fall 2026.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-17 and the Technical Specifications Date of Issuance: March 25, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.25 15:09:34 -04'00'
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION ATTACHMENT AMENDMENT NO. 323 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-35 AMENDMENT NO. 319 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-52 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 Renewed Facility Operating License Nos. NPF-35 and NPF-52 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 Appendix A to Renewed Facility Operating License Nos. NPF-35 and NPF-52 Replace the following pages of the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert TS 4.0-1 TS 4.0-1 TS 5.6-3 TS 5.6-3 TS 5.6-4 TS 5.6-4 TS 5.6-5 Renewed License No. NPF-35 Amendment No. 323 (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 323, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.
(3)
Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),
following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
(4)
Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.
(5)
Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),
the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Renewed License No. NPF-52 Amendment No. 319 (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.
(3)
Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),
following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
(4)
Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.
(5)
Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),
the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Design Features 4.0 (continued)
Catawba Units 1 and 2 4.0-1 Amendment Nos. 323/319 4.0 DESIGN FEATURES 4.1 Site Location Catawba Nuclear Station is located in the north central portion of South Carolina approximately six miles north of Rock Hill and adjacent to Lake Wylie. The station center is located at latitude 35 degrees, 3 minutes, 5 seconds north and longitude 81 degrees, 4 minutes, 10 seconds west. The corresponding Universal Transverse Mercator Coordinates are E 493, 660 and N 3, 878, 558, zone 17.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, Optimized ZIRLO', or AXIOM (Unit 1 only) clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of ZIRLO, Optimized ZIRLO', AXIOM (Unit 1 only), zirconium alloy, or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium and boron carbide as approved by the NRC.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
Reporting Requirements 5.6 (continued)
Catawba Units 1 and 2 5.6-3 Amendment Nos. 323/319 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY" (W Proprietary).
- 2.
Deleted.
- 3.
Deleted.
- 4.
DPC-NE-2011-P-A, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors" (DPC Proprietary).
- 5.
DPC-NE-3001-P-A, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology" (DPC Proprietary).
- 6.
DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design."
- 7.
DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology."
- 8.
DPC-NE-3000-P-A, "Thermal-Hydraulic Transient Analysis Methodology" (DPC Proprietary).
- 9.
DPC-NE-1004-A, "Design Methodology Using CASMO-3/SIMULATE-3P."
- 10.
DPC-NE-2004-P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01" (DPC Proprietary).
- 11.
DPC-NE-2005-P-A, "Thermal Hydraulic Statistical Core Design Methodology" (DPC Proprietary).
- 12.
DPC-NE-2008-P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3" (DPC Proprietary).
Reporting Requirements 5.6 (continued)
Catawba Units 1 and 2 5.6-4 Amendment Nos. 323/319 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 13.
WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 14.
DPC-NE-2009-P-A, Westinghouse Fuel Transition Report (DPC Proprietary).
- 15.
WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 16.
DPC-NE-1005P-A, Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, (DPC Proprietary).
- 17.
Deleted.
- 18.
DPC-NE-1007-PA, Conditional Exemption of the EOC MTC Measurement Methodology (Duke and Westinghouse Proprietary).
- 19.
WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995 (Westinghouse Proprietary).
- 20.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO', July 2006 (Westinghouse Proprietary).
- 21.
WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (Westinghouse Proprietary). [For use by Unit 1 only.]
- 22.
WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel (Westinghouse Proprietary). [For use by Unit 1 only.]
The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e., report number, title, revision number, report date or NRC SER date, and any supplements).
Reporting Requirements 5.6 (continued)
Catawba Units 1 and 2 5.6-5 Amendment Nos. 323/319 5.6 Reporting Requirements (continued)
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Not used.
5.6.7 PAM Report When a report is required by LCO 3.3.3, "Post Accident Monitoring (PAM)
Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION ATTACHMENT AMENDMENT NO. 333 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-9 AMENDMENT NO. 312 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-17 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 Renewed Facility Operating License Nos. NPF-9 and NPF-17 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert NPF-9, page 3 NPF-9, page 3 NPF-17, page 3 NPF-17, page 3 Appendix A to Renewed Facility Operating License Nos. NPF-9 and NPF-17 Replace the following pages of the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert TS 4.0-1 TS 4.0-1 TS 5.6-3 TS 5.6-3 TS 5.6-4 TS 5.6-4 TS 5.6-5
Renewed License No. NPF-9 Amendment No. 333 (4)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6)
Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 333, are hereby incorporated into this renewed operating license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Renewed License No. NPF-17 Amendment No. 312 (4)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6)
Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 312, are hereby incorporated into this renewed operating license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Design Features 4.0 McGuire Units 1 and 2 4.0-1 Amendment Nos. 333/312 4.0 DESIGN FEATURES 4.1 Site Location The McGuire Nuclear Station site is located at latitude 35 degrees, 25 minutes, 59 seconds north and longitude 80 degrees, 56 minutes, 55 seconds west. The Universal Transverse Mercator Grid Coordinates are E 504, 669, 256, and N 3, 920, 870, 471.
The site is in northwestern Mecklenburg County, North Carolina, 17 miles north-northwest of Charlotte, North Carolina.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, Optimized ZIRLOTM, or AXIOM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of ZIRLO, Optimized ZIRLOTM, AXIOM, zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium (Unit 1) silver indium cadmium and boron carbide (Unit 2) as approved by the NRC.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a.
Fuel assemblies having a maximum nominal U-235 enrichment of 5.00 weight percent;
- b.
keff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
- c.
keff < 0.95 if fully flooded with water borated to 800 ppm, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
Reporting Requirements 5.6 (continued)
McGuire Units 1 and 2 5.6-3 Amendment Nos. 333/312 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY, (W Proprietary).
- 2.
Deleted.
- 3.
Deleted.
- 4.
DPC-NE-2011PA, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).
- 5.
DPC-NE-3001PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).
- 6.
DPC-NF-2010A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design ".
- 7.
DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology.
- 8.
DPC-NE-3000PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Proprietary).
- 9.
DPC-NE-1004A, "Nuclear Design Methodology Using CASMO -
3/SIMULATE-3 P ".
- 10.
DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, (DPC Proprietary).
- 11.
DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Proprietary).
- 12.
DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3," (DPC Proprietary).
- 13.
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code, (W Proprietary) [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
Reporting Requirements 5.6 (continued)
McGuire Units 1 and 2 5.6-4 Amendment Nos. 333/312 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 14.
DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report, (DPC Proprietary).
- 15.
WCAP-12945-P-A, Volume 1 and Volumes 2-5, " Code Qualification Document for Best-Estimate Loss of Coolant Analysis," (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 16.
DPC-NE-1005P-A, Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, (DPC Proprietary).
- 17.
DPC-NE-1007-PA, Conditional Exemption of the EOC MTC Measurement Methodology (Duke and Westinghouse Proprietary).
- 18.
WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995. (Westinghouse Proprietary).
- 19.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," July 2006. (Westinghouse Proprietary).
- 20.
WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (Westinghouse Proprietary).
- 21.
WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel (Westinghouse Proprietary).
The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e., report number, title, revision number, report date or NRC SER date, and any supplements).
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Deleted
Reporting Requirements 5.6 (continued)
McGuire Units 1 and 2 5.6-5 Amendment Nos. 333/312 5.6 Reporting Requirements 5.6.7 PAM Report When a report is required by LCO 3.3.3, "Post Accident Monitoring (PAM)
Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG;
- b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
- c.
For each degradation mechanism found:
- 1. The nondestructive examination techniques utilized;
- 2. The location, orientation (if linear), measured size (if available),
and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
- 3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
- 4. The number of tubes plugged during the inspection outage.
- d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
- e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and
- f.
The results of any SG secondary side inspections.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO CATAWBA NUCLEAR STATION, UNITS 1 AND 2 AMENDMENT NO. 323 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-35 AMENDMENT NO. 319 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-52 AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 AMENDMENT NO. 333 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-9 AMENDMENT NO. 312 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-17 DUKE ENERGY CAROLINAS, LLC DOCKET NOS. 50-413, 50-414, 50-369, AND 50-370
1.0 INTRODUCTION
By letter RA-25-0014 dated August 18, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25230A072, as supplemented by letter RA-25-0280 dated January 8, 2026 (ML26008A039), Duke Energy Carolinas, LLC (Duke Energy, or the licensee) submitted a license amendment request (LAR) to Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and NPF-17 for the Catawba Nuclear Station (CNS), Units 1 and 2, and the McGuire Nuclear Station (MNS), Units 1 and 2, respectively. The proposed changes would approve the use of the Westinghouse fuel cladding alloy designated as AXIOM and adoption of the Westinghouse Full Spectrum Loss of Coolant Accident Methodology at CNS, Unit 1, and MNS, Units 1 and 2. Conforming changes are made to the TS for CNS, Unit 2, but the licensee plans to submit a separate LAR to use AXIOM fuel rod cladding and adopt the FULL SPECTRUM' LOCA (FSLOCA') Methodology for CNS, Unit 2.
Specifically, the amendments would revise Catawbas and McGuires Technical Specifications (TSs) to include Westinghouse topical report WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology), (ML17277A130) to the list of NRC-approved analytical methods used to determine the core operating limits in TS 5.6.5, CORE OPERATING LIMITS REPORT. The
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION amendments also annotate select legacy LOCA methods and the deletion of others to restrict their future use and allow for a staggered implementation during refueling outages at each unit.
The amendments also revise TSs 4.2.1, Fuel Assemblies, to permit the use of the Westinghouse fuel cladding alloy designated as AXIOM. Specifically, the proposed amendment requests a revision to the TSs to update the description of fuel assemblies specified in TS 4.2.1, Fuel Assemblies, and add the Westinghouse topical report WCAP-18546-P-A, Westinghouse AXIOM Cladding for use in Pressurized Water Reactor Fuel (ML23089A063) to the referenced NRC-approved analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding.
In order to support the TSs change, and pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.12, Specific exemptions, in Enclosure 5 to the LAR, Duke Energy requested an exemption from certain requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems [ECCSs] for light-water nuclear power reactors, for CNS, Unit 1, and MNS, Units 1 and 2. The exemption request relates solely to the specific type of cladding material described in the regulation for use in light-water reactors, as the AXIOM fuel rod cladding material is not an explicitly identified fuel rod cladding material listed in 10 CFR 50.46. On March 10, 2026, the NRC issued the exemption (ML26014A011).
From November 7, 2025, through January 22, 2026, the U.S. Nuclear Regulatory Commission (NRC) staff conducted a regulatory audit to support its review of the amendment request, as discussed in the staffs audit plan dated November 7, 2025 (ML25273A179), and audit summary dated February 2, 2026 (ML26027A080).
1.1 Proposed Changes`
The licensee requested to modify TS 4.2.1, Fuel Assemblies, and TS 5.6.5 CORE OPERATING LIMITS REPORT, to allow the use of AXIOM as an approved fuel rod cladding and FSLOCATM analytical methodology.
The specific proposed changes to TSs are shown below where bold red text indicates additions and bold strikeout text indicates deletions.
For CNS, Units 1 and 2 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, or Optimized ZIRLOTM, or AXIOM (Unit 1 only) clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.* Limited substitutions of ZIRLO, Optimized ZIRLOTM, AXIOM (Unit 1 only),
zirconium alloy, or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION A maximum of four lead assemblies containing mixed oxide fuel and M5TM cladding may be inserted into the Unit 1 or Unit 2 reactor core.
5.6.5 COLR
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY (W Proprietary).
- 2.
WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE (W Proprietary).
- 3.
BAW-10168-P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants (B&W Proprietary).
- 4.
DPC-NE-2011-P-A, Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors (DPC Proprietary).
- 5.
DPC-NE-3001-P-A, Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology (DPC Proprietary).
- 6.
DPC-NF-2010-A, Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design.
- 7.
DPC-NE-3002-A, FSAR Chapter 15 System Transient Analysis Methodology.
- 8.
DPC-NE-3000-P-A, Thermal-Hydraulic Transient Analysis Methodology (DPC Proprietary).
- 9.
DPC-NE-1004-A, Design Methodology Using CASMO-3/SIMULATE-3P.
- 10.
DPC-NE-2004-P-A, Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01 (DPC Proprietary).
- 11.
DPC-NE-2005-P-A, Thermal Hydraulic Statistical Core Design Methodology (DPC Proprietary).
- 12.
DPC-NE-2008-P-A, Fuel Mechanical Reload Analysis Methodology Using TACO3 (DPC Proprietary).
- 13.
WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 14.
DPC-NE-2009-P-A, Westinghouse Fuel Transition Report (DPC Proprietary).
- 15.
WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 16.
DPC-NE-1005P-A, Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, (DPC Proprietary).
- 17.
BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, (Framatome ANP Proprietary). Deleted.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 18.
DPC-NE-1007-PA, Conditional Exemption of the EOC MTC Measurement Methodology (Duke and Westinghouse Proprietary).
- 19.
WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995 (Westinghouse Proprietary).
- 20.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO',
July 2006 (Westinghouse Proprietary).
- 21.
WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)
(Westinghouse Proprietary). [For use by Unit 1 only.]
- 22.
WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel (Westinghouse Proprietary). [For use by Unit 1 only.]
For MNS, Units 1 and 2 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, or Optimized ZIRLOTM, or AXIOM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.
Limited substitutions of ZIRLO, Optimized ZIRLOTM, AXIOM, zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
5.6.5 COLR
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 22. WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY (W Proprietary).
- 23.
WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE, (W Proprietary).
- 24.
BAW-10168P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B& W Proprietary).
- 25.
DPC-NE-2011PA, Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, (DPC Proprietary).
- 26.
DPC-NE-3001PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology, (DPC Proprietary).
- 27.
DPC-NF-2010A, Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 28.
DPC-NE-3002A, FSAR Chapter 15 System Transient Analysis Methodology.
- 29.
DPC-NE-3000PA, Thermal-Hydraulic Transient Analysis Methodology,
(DPC Proprietary).
- 30.
DPC-NE-1004A, Nuclear Design Methodology Using CASMO - 3/SIMULATE-3 P.
- 31.
DPC-NE-2004P-A, Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, (DPC Proprietary).
- 32.
DPC-NE-2005P-A, Thermal Hydraulic Statistical Core Design Methodology, (DPC Proprietary).
- 33.
DPC-NE-2008P-A, Fuel Mechanical Reload Analysis Methodology Using TACO3, (DPC Proprietary).
- 34. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code, (W Proprietary) [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 35.
DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report, (DPC Proprietary).
- 36. WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis, (W Proprietary). [Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.]
- 37.
DPC-NE-1005P-A, Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, (DPC Proprietary).
- 38.
DPC-NE-1007-PA, Conditional Exemption of the EOC MTC Measurement Methodology (Duke and Westinghouse Proprietary).
- 39.
WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995. (Westinghouse Proprietary).
- 40.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO',
July 2006. (Westinghouse Proprietary).
- 41.
WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)
(Westinghouse Proprietary).
- 42.
WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel (Westinghouse Proprietary).
2.0 REGULATORY AND GUIDANCE REQUIREMENTS 2.1 Regulatory Requirements The NRCs regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications.
This regulation requires that TSs include: (1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The regulation in 10 CFR 50.36(c)(4), Design features states:
Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.
The regulation in 10 CFR 50.36(c)(5), Administrative controls states, in part, that:
Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The regulation in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Section 50.46(a)(1)(i) states that, Each pressurized or boiling water reactormust be provided with an acceptable ECCS evaluation model [EM] and must be calculated for a number of postulated loss-of-coolant accident sizes, locations, and other properties sufficient to provide assurance that the most severe loss-of-coolant accidents are calculatedand that uncertainty must be accounted for, so that, when the calculated ECCS performance is compared to the criteria set forth in paragraph (b) of this Section there is a high level of probability that the criteria would not be exceeded.
The regulations in 10 CFR 50.46(b) require, in part, that during a LOCA event, the following criteria are satisfied:
(1)
Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200° F.
(2)
Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
(3)
Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4)
Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5)
Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, (hereinafter referred to as GDC), establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The GDC that are relevant to this LAR include:
GDC 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
GDC 15, Reactor coolant system design, requires that the reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
GDC 35, Emergency core cooling, requires, in part, that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal water reaction is limited to negligible amounts.
2.2 Guidance Documents NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), Section 4.2, Revision 3, Fuel System Design, March 2007 (ML070740002)
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, March 2007 (ML070550016).
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the regulations and guidance discussed in Section 2.0 of this safety evaluation.
The NRC staff reviewed the proposed changes to verify that the AXIOM fuel cladding material is appliable and that all limitations and conditions are met for use in CNS, Unit 1, and MNS, Units 1 and 2.
3.1 Evaluation of TS Changes 3.1.1 Evaluation of Design Feature TS 4.2.1 Changes The licensee proposed addition of the Westinghouse AXIOM alloy to the list of materials that may be used as the fuel rod cladding in CNS, Unit 1, and MNS, Units 1 and 2, fuel assemblies.
The proposed change for CNS includes a parenthetical note after each mention of AXIOM that limits the TS applicability to (Unit 1 only). The licensee plans to submit a separate LAR for CNS, Unit 2, to use AXIOM fuel rod cladding and adopt FSLOCA'. In its letter dated January 8, 2026, the licensee proposed to remove the current note in TS 4.2.1, applicable only
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION to CNS, regarding the maximum number of lead assemblies containing mixed oxide fuel and M5 cladding materials. The licensee proposed deleting the note because it is no longer considering installation of these assemblies. In its submittal, the licensee stated that it is using the approved topical reports associated with both AXIOM (WCAP-18546-P-A, Revision 0), and FSLOCA' (WCAP-16996-P-A, Revision 1) methodologies. Conformance with these two topical reports is addressed below in Sections 3.2 and 3.3, respectively.
In addition, the licensee has submitted a request in the LARs Enclosure 5 for the required exemption from portions of 10 CFR 50.46. On March 10, 2026, the NRC staff issued the exemption (ML26014A011).
Based on the above, the NRC staff found that the proposed addition of AXIOM to the list of materials that may be used as fuel cladding material in TS 4.2.1, Fuel Assemblies, to be consistent with the NRC-approved topical reports and is, therefore, acceptable. In addition, the NRC staff found that the proposed deletion of the note regarding Catawba lead assemblies is no longer applicable to the design and is, therefore, acceptable. Therefore, the NRC staff finds that with the proposed changes 10 CFR 50.36(c)(4) will continue to be met as TS 4.2.1 continues to include design features of the facility, which, if altered or modified, would have a significant effect on safety.
3.1.2 Evaluation of the TS 5.6.5 Changes The licensee proposed an administrative change to clean up the list of NRC-approved analytical methods provided in CNS and MNS TS 5.6.5 by deleting two legacy LOCA methodologies. The licensee stated that these methods have not been used since the transition from Framatome Mark-BW fuel (BAW-10168-P-A) to Westinghouse Robust Fuel Assembly (RFA) fuel in 1998, and implementation of Westinghouse Best Estimate Large Break Loss of Coolant Accident (LBLOCA) analysis in 2000 which replaced the BASH CODE LBLOCA method (WCAP-10266-P-A). Since the methods are not used, the NRC staff found that the deletion of WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE, and BAW-10168-P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, from CNS and MNS TS 5.6.5 is administrative in nature and is, therefore, acceptable.
The licensee also proposed to delete a fuel rod methodology, BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, that was used to support Framatome Mixed Oxide (MOX) fuel lead test assemblies. The licensee stated that this fuel is no longer under consideration for use as a batch feed fuel. Therefore, the NRC staff notes that fuel supported by this analytical methods is not present in the reactor core.
The licensee proposed another administrative change to annotate two LOCA methodologies to indicate when they will no longer be used to establish core operating limits. These are WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code, and WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis. The licensee stated that these will no longer be used once all of the Optimized ZIRLO' cladding is discharged from the core. As stated in its supplement January 8, 2026, the annotation for these two TRs is as follows:
Shall not be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff found the proposed annotation to be acceptable as those methodologies are being replaced with the FSLOCA' EM.
The licensee proposed changes to CNS and MNS TS 5.6.5 to add Westinghouse topical reports WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), and WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, to the list of NRC-approved analytical methods. The NRC staff considers the addition of the topical reports for AXIOM (WCAP-18546-P-A) and FSLOCA' (WCAP-16996-P-A) to the COLR Reference list in CNS and MNS TS 5.6.5.b acceptable as the methodologies are NRC-approved and found acceptable for CNS, Unit 1, and MNS, Units 1 and 2, as discussed in Sections 3.2 and 3.3 of this safety evaluation. Based on the above, the NRC staff finds that the proposed changes continue to provide sufficient administrative controls in accordance with 10 CFR 50.36(c)(5).
3.2 Evaluation of AXIOM Topical Report Applicability As discussed in the AXIOM topical report, AXIOM cladding is a niobium-bearing zirconium alloy like the ZIRLO' alloy, with reduced tin content to increase corrosion resistance like the Optimized ZIRLO' alloy, and with the addition of other alloying elements including vanadium and copper to improve specific properties like hydrogen pickup. The AXIOM alloy has been processed to be in the partially recrystallized annealed condition similar to the Optimized ZIRLO cladding to compensate for the creep strength loss caused by the reduced tin content. The topical report describes in detail how the properties and performance of AXIOM cladding are incorporated into existing NRC-approved analytical methods for use in plant-specific safety analyses.
3.2.1 AXIOM Limitations & Conditions The safety evaluation for WCAP-18546-P, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor [PWR] Fuel, Revision 0, March 2023 (ML23089A066) contains limitations and conditions that must be met in order to be implemented. Each limitation and condition is presented below, followed by a description of how the licensee addressed the limitation and condition, and the associated NRC staff findings.
Reactor and Fuel Assembly Designs AXIOM cladding must be used with the NRC-approved PWR designs AXIOM cladding must be used with the NRC-approved Westinghouse and CE fuel designs with corresponding pellet and assembly dimensions AXIOM cladding must be used with the NRC-approved fuel materials and pellet coatings or additives (e.g., advanced doped pellet technology (ADOPT'), Integral Fuel Burnable Absorber (IFBA), gadolinium)
Licensee Compliance In Enclosure 1, Section 3.0, Technical Evaluation, of its submittal dated August 18, 2025, the licensee states, in part, that:
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Both MNS units are PWRs that are currently licensed for operation by the NRC per renewed operating licenses NPF-9 and NPF-17, respectively. Similarly, CNS, Unit 1, is a PWR that is currently licensed for operation by the NRC per renewed operating license NPF-35. Duke Energys use of AXIOM cladding will not challenge this limitation related to PWR design.
Both MNS and CNS, Unit 1, utilize the 17 x 17 Westinghouse RFA [Robust Fuel Assembly] design, an NRC-approved fuel design. Duke Energys use of AXIOM cladding will not challenge this limitation related to fuel design.
The limitation for use of NRC-approved fuel materials and pellet coatings or additives will be controlled through the fuel procurement process or validated by cycle-specific engineering analyses. Duke Energys use of AXIOM cladding will not challenge this limitation related to fuel materials and pellet coatings or additives.
Based on its review of the above information, the NRC staff found that the limitations and conditions relating to the reactor and fuel assembly design are met.
Fuel Limitations Currently fuel burnup shall be limited to 62 GWd/MTU peak rod average for all cladding types, however, fuel rod burnup (( )) may be allowed once additional information specific to burnup to (( )) is submitted and approved by the NRC Best Estimate Oxide Thickness < 100 m Best Estimate HPU (Hydrogen Pickup) (( ))
Licensee Compliance In Enclosure 1, Section 3.0, Technical Evaluation, of its submittal dated August 18, 2025, the licensee states, in part, that:
While there is the potential for an increased limit once additional information is submitted and approved by the NRC, Duke Energys use of AXIOM cladding for MNS and CNS, Unit 1, will apply a peak rod average burnup limit of 62 GWd/MTU. This limitation will be validated by cycle-specific engineering analyses As provided in Section 5.1.1 of WCAP-18546-P-A, Operating PWRS, the measured maximum oxide thickness of the AXIOM alloys are less than 50 m for a burnup of close to 75 GWd/MTU. The best estimate oxide thickness will be less than the allowed 100 m for a peak rod average burnup of 62 GWd/MTU. Furthermore, this limitation will be validated by cycle-specific engineering analyses As shown in Section 5.2 of WCAP-18546-P-A, Hydrogen Pickup, the overall maximum hydrogen content for AXIOM is significantly less than the value specified in the Topical Reports safety evaluation, Section 4.0 Limitation and Conditions bullet no. 6, which thereby demonstrates AXIOMs low HPU. This limitation will be validated by cycle-specific engineering analyses to ensure compliance with this condition and limitation.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Based on the above, the NRC staff found that the limitations and conditions relating to specific fuel limitations are met.
3.2.2 AXIOM Conclusion Based on its review of the technical information provided by the licensee, which documented compliance to the limitations and conditions for the AXIOM topical report, the NRC staff finds the use of WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, applicable to CNS, Unit 1, and MNS, Units 1 and 2, acceptable for addition to TS 5.6.5.b. as this continues to meet the administrative controls criteria and meets 10 CFR 50.36(c)(5).
3.3 Evaluation of FSLOCA' Topical Report Applicability The NRC staff reviewed the proposed changes to verify that the new LOCA methodology is an approved NRC code and that all limitations and conditions are met, that the licensee appropriately applied the LOCA EM to CNS Unit 1 and MNS Units 1 and 2, and that the results meet the acceptance criteria of 10 CFR 50.46(b)(1) through (4).
As described in WCAP-16996-P-A, Revision 1, the purpose of the FSLOCATM Methodology EM is to build on the ASTRUM EM, by extending the applicability of the WCOBRA/TRAC code to include the treatment of small break LOCA (SBLOCA) and intermediate break LOCA (IBLOCA) scenarios. The term Full Spectrum specifies that the new EM is intended to resolve the full spectrum of LOCA scenarios that result from a postulated break in the cold leg of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA methodology include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture with a break flow area equal to two times the pipe area.
3.3.1 FSLOCA' Limitations & Conditions The safety evaluation for WCAP-16996-P-A, Revision 1, November 2016 (ML17277A132) contains 15 limitations and conditions that must be met in order to implement the NRC-approved FSLOCA EM.
Each limitation and condition is presented below, followed by a description of how the licensee addressed the limitation and condition, and the associated NRC staff finding.
Limitation and Condition 1 - FSLOCATM EM Applicability with Regard to LOCA Transient Phases In Section 5.0 Limitations and Conditions, of the safety evaluation, Table 22, Limitations and Conditions Based on the Technical Evaluation of the Updated FSLOCATM EM Documented in WCAP-16996-P/WCAP-16996-NP, Volume I, II, and III, Revision 1, states, in part, that The FSLOCATM EM applicability for performing PWR LOCA analyses is defined in terms of applicable accident transient phases so that the FSLOCA' EM cannot be applied for analyzing the long-term core cooling phase of LOCA transients for the purpose of demonstrating compliance with the long-term core cooling requirement set forth in 10 CFR 50.46(b)(5). This limitation specifically addresses the condition that the
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION FSLOCATM EM does not treat boric acid precipitation and therefore lacks capabilities to address adequately post-LOCA long-term core cooling.
The numerical approximations to advection and diffusion in the WCOBRA/TRAC-TF2 code conservation equations have neither been validated nor shown to successfully track the movement of high concentrations of boric acid between the vertical and radial cells with the vessel volumes.
Licensee Compliance In Section 2.3 of Enclosure 2, Compliance with FSLOCA EM Limitations and Conditions, of its submittal dated August 18, 2025, the licensee states that:
The analysis for McGuire Units 1 and 2 and Catawba Unit 1 with the FSLOCA EM is only being used to demonstrate compliance with the applicable ECCS acceptance criteria discussed in the LARs, Enclosure 2, Section 6.0, Compliance with Applicable ECCS Acceptance Criteria, and is not being used to demonstrate compliance with 10 CFR 50.46(b)(5).
Given that the licensee is not using the FSLOCA EM to demonstrate compliance with 10 CFR 50.46(b)(5), the NRC staff found that the licensee has met the requirements of Limitation and Condition 1.
Limitation and Condition 2 - FSLOCATM EM Applicability with Regard to Type of PWR Plants In Section 5.0, Table 22 of the safety evaluation states, in part, that The FSLOCATM EM applicability for performing PWR LOCA analyses is defined in terms of applicable types of PWR plants so that the EM can be applied for LOCA analyses of Westinghouse-designed three-loop and four-loop PWR plants with cold-side emergency core cooling injection, only. Plant-specific applications will generally be considered acceptable if they follow the requirements pertinent to FSLOCA described in [ the safety evaluation for FSLOCA ADAMS Accession No. ML17277A132] and comply and meet the NRC limitations and conditions in this table (where the later document supersedes the earlier document when differences exist). Plant-specific licensing actions referencing FSLOCA analyses should include a statement summarizing the extent to which the FSLOCA methods and modeling were followed, and justification for any departures.
Should NRC staff review determine that absolute adherence to the modeling guidelines is inappropriate for a specific plant, additional information may be requested using the Request for Additional Information (RAI) process.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that; McGuire Units 1 and 2 and Catawba Unit 1 are Westinghouse-designed 4-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The analysis for McGuire Units 1 and 2 and Catawba Unit 1 utilized the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION pursuant to 10 CFR 50.46 in References 10 through 16 [ Enclosure 4, Section 8.0, References] and the changes for applications of AXIOM cladding, as described in Reference 3 [Enclosure 4, Section 8.0].
The analysis was performed with a code version which incorporated the changes and error corrections described in References 10 through 16 [Enclosure 4, Section 8.0],
except for the error in the steam/fission gas specific heat calculation described in Reference 16 [Enclosure 4, Section 8.0] This error was found to have a negligible impact on analysis results with the FSLOCA EM, leading to an estimated PCT [peak cladding temperature] impact of 0°F, as described in Reference 16 [Enclosure 4, Section 8.0].
Given that CNS, Unit 1, and MNS, Units 1 and 2, are Westinghouse-designed 4-loop PWRs with cold-side injection, the NRC staff found that the FSLOCA EM is applicable. In addition, the NRC staff found that the licensee has appropriately applied the FSLOCA EM.
The NRC staff reviewed the changes in the Westinghouse 10 CFR 50.46 Annual Notification and Reporting letters (LAR, Enclosure 4, References 10 through 16) specific to the FSLOCA EM. The various changes/corrections include items in several broad categories including general code maintenance, error corrections and improvements. General code maintenance includes items such as improving input diagnostic checks, enhancing the code output, optimizing active coding, and eliminating inactive coding. The error corrections were all evaluated by Westinghouse and summarized in the LARs Enclosure 4, References 10 through 16 and all were found to have an estimated peak cladding temperature impact of 0°F. The several improvements included an enhancement to the pump momentum equation at low pump speed, updates to the kinetics and decay heat model for higher burnup fuel and adjustments to the fuel rod radial noding to be consistent with the PAD5 fuel rod performance analysis and design model code.
The one error correction that was not addressed in the Catawba/McGuire FSLOCA analysis was the steam/fission gas specific heat calculation error. An incorrect specific heat value results in the superheated steam thermal conductivity being under-predicted at the burst node after rupture is predicted. This leads to a reduction in the gap conductance and over-prediction of the fuel average temperatures at the rod burst location after burst is predicted. Westinghouse qualitatively evaluated the error and found that it leads to an estimated peak cladding temperature impact of 0°F.
During previous NRC staff reviews1 of applications of the FSLOCA EM, an error was reported related to the gamma redistribution multiplier on the hot rod and hot assembly power. This error was not reported in the Westinghouse 10 CFR 50.46 annual reports. During this LARs regulatory audit, staff inquired about this error and it was noted that the error only impacted certain plants and therefore, wasnt included in the annual reports. In addition, in the licensees letter dated January 8, 2026, the licensee confirmed that the current analysis for CNS, Unit 1, and MNS, Units 1 and 2, was performed with a code version that incorporated the gamma redistribution multiplier error correction.
1 Diablo Canyon - ADAMS Accession No. ML19266A657 and Surry - ADAMS Accession No. ML19309D196.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Based on its review of the above information, the NRC staff found that the changes/corrections to the FSLOCA EM are appropriate and acceptable. Therefore, the NRC staff found that the licensee met the requirements of Limitation and Condition 2.
Limitation and Condition 3 - FSLOCATM EM Applicability for Containment Pressure Modeling In Section 5.0, Table 22 of the safety evaluation states, in part, that:
The coupled WCOBRA/TRAC-TF2 and COCO codes or standalone LOTIC2 code will be applied to calculate the containment backpressure in PWR LOCA analyses for Region II so that a conservatively low, although not explicitly bounded, containment pressure will be predicted and used. For this purpose, the input to the COCO model and its prediction results will be based on appropriate plant-specific containment design parameters and initial conditions and will simulate accordingly engineered safety features and installed systems capable of affecting the containment pressure including their actuation, performance, and associated processes. The following specific limitations will apply for Region II analyses using the FSLOCATM EM: (1) an acceptable plant-specific initial containment temperature will be determined based on input from the utility for the purpose of modeling the containment pressure response with COCO or LOTIC2; and (2) unqualified or indeterminate coatings throughout containment and qualified coatings within the break jet zone-of-influence will not be credited for the purpose of modeling the containment pressure response using COCO or LOTIC2 consistent with the bounding treatment of this parameter (conservatively low containment pressure). Please see LTR-NRC-15-102, Revision 2 (pages P-7 to P-10) for containment modeling.
Licensee Compliance In Section 2.3 of Enclosure 2, the licensee states, in part, that:
The containment pressure calculation for the McGuire Units 1 and 2 and Catawba Unit 1 analysis was performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A plant-specific initial temperature associated with normal full-power operating conditions (applicable to all three units) was modeled, and no coatings were credited on any of the containment structures.
The NRC staff found that the licensee used the NRC-approved LOTIC2 methodology for the Region II containment pressure calculation with appropriate design parameters and conditions.
Therefore, the NRC staff found that the licensee has met the requirements of Limitation and Condition 3.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation and Condition 4 - Decay Heat Modeling in FSLOCATM EM Applications In Section 5.0, Table 22 of the safety evaluation states, in part, that:
As implemented by Westinghouse and found acceptable from the review of the decay heat model in the FSLOCATM EM, the following conditions will apply with regard to decay heat modeling and sampling in PWR LOCA analyses for Region I and Region II:
(1) decay heat uncertainty will be (( )) in uncertainty analyses for both Region I and Region II according to Table 29-4 in WCAP-16996-P/WCAP-16996-NP, Revision 1, Volume III, Section 29; (2) the FSLOCATM EM cannot be applied for transient time longer than 10,000 seconds following shutdown unless the decay heat model is shown to be acceptable for the analyzed core conditions. The latter limitation is (( )) The sampled value of the decay heat uncertainty multiplier, DECAY_HT, reported in units of and absolute units, as applied for the limiting runs in Region I and Region II in the plant-specific analysis as part of a License Amendment Request submittal, will be provided as part of the submittal.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that; Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was (( )) for the McGuire Units 1 and 2 and Catawba Unit 1 analysis. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip.
The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Table 16 [of Enclosure 2 to the LAR].
The NRC staff found that the licensee appropriately modeled decay heat per the limitation and condition and reported the resulting sampled values in units of sigma and absolute units for the limiting cases. Therefore, the NRC staff found that the licensee has met the requirements of Limitation and Condition 4.
Limitation and Condition 5 - Fuel Burnup Limits in FSLOCATM EM Applications In Section 5.0, Table 22 of the safety evaluation states, in part, that:
The maximum assembly average burnup will be limited to (( )) and the maximum peak rod length-average burnup will be limited to (( )) within the FSLOCATM EM. See WCAP-16996-P, Revision 1, Section 32.4, Methodology Limitations, page 32-21.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
The maximum analyzed assembly and rod length-average burnup were less than or equal to (( )) respectively, for McGuire Units 1 and 2 and Catawba Unit 1.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Based on the above, the NRC staff found that the licensee has met the requirements of Limitation and Condition 5.
Limitation and Condition 6 - WCOBRA/TRAC-TF2 Interface with PAD 5.0 in the FSLOCATM EM In Section 5.0, Table 22 of the safety evaluation states, in part, that:
In the FSLOCATM EM applications for PWR LOCA analyses, the latest version of an NRC-approved version of the latest fuel performance code that is applicable for the LOCA analysis will be used to initialize the fuel rod initial conditions. If the PAD 5.0 code is the latest approved version for fuel performance LOCA evaluations, then this version will be used to interface with WCOBRA/TRAC-TF2. The fuel performance code utilized shall be used to initialize WCOBRA/TRAC-TF2 using appropriate calculative methods to maximize the initial fuel stored energy and gap pin pressure, as well as adhere to any restrictions and limitations that resulted from the staff review and acceptance. The fuel performance code calculative methods should therefore exercise those modeling techniques approved by the staff for initializing WCOBRA/TRAC-TF2 for LOCA evaluations. The fuel performance code shall also include the effects of fuel thermal conductivity degradation and its attendant effects on fuel rod behavior for application to the WCOBRA/TRAC-TF2 code.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
PAD5 fuel performance data were utilized in the McGuire Units 1 and 2 and Catawba Unit 1 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 2
[Westinghouse Performance Analysis and Design Model (PAD5), WCAP-17642-P-A, Revision 1, November 2017], and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 2 [Westinghouse Performance Analysis and Design Model (PAD5), WCAP-17642-P-A, Revision 1, November 2017].
Given that the licensee used the latest NRC-approved fuel performance code (i.e., PAD5) and used appropriate conservative inputs, the NRC staff found that the licensee has met the requirements of Limitation and Condition 6.
Limitation and Condition 7 - Interfacial Drag Uncertainty in FSLOCATM EM Region I Analyses In Section 5.0, Table 22 of the safety evaluation states, in part, that:
As implemented by Westinghouse and found appropriate based on the review of the two-phase interfacial drag model of the 3D VESSEL module in WCOBRA/TRAC-TF2 and its assessment, the interfacial drag multiplier, YDRAG, applied to the small bubble, small-to-large bubble, and churn-turbulent flow regimes of the Cold Wall two-phase flow map and to the Hot Wall two-phase flow map interfacial drag will be (( ))
established for YDRAG in the FSLOCATM EM as described in WCAP-16996-P/WCAP-16996-NP, Revision 1, Section 13.4 and Section 29.1.5 as lower interfacial drag reduces the two-phase mixture thus promoting core uncovery. This (( ))
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The proprietary SEs Table 22: Limitations and Conditions Based on the Technical Evaluation of the Updated FSLOCATM EM Documented in WCAP-16996-P/WCAP-16996-NP, Volume I, II, and III, Revision 1 (Continued) further states that this item is; The comprehensive list of (( )) is given in Table 29.2.3-1 of WCAP-16996-P, Revision 1 (see page 29-52).
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was
(( )) for the McGuire Units 1 and 2 and Catawba Unit 1 Region I analysis.
The NRC staff found that the licensee appropriately used the specified interfacial drag uncertainty parameter as noted above and, therefore, meets the requirement of Limitation and Condition 7.
Limitation and Condition 8 - Biased Uncertainty Contributors in FSLOCATM EM Region I Analyses In Section 5.0, Table 22 of the safety evaluation states, in part, that:
As implemented by Westinghouse and found acceptable from the review of the corresponding WCOBRA/TRAC-TF2 models, certain uncertainty contributors will be (( ))
for Region I analyses with the FSLOCATM EM according to Table 29.2.3-1 and Table 29-2 in WCAP-16996-P/WCAP-16996-NP, Revision 1, Volume III, Section 29.2.3.
Specifically, the (( )) as established in the FSLOCATM EM and described in WCAP-16996-P, Revision 1, Section 17.2.3 and Section 29.1.6 for KCOSI and in Section 4.4.5 and Section 29.1.7 for HS_SLUG. Lower condensation heat transfer in the cold legs may influence depressurization rate during an SBLOCA boil-off period. A higher transition boundary delays transition to non-stratified flow thus increasing residual liquid in the loop seal regions and decreasing vapor venting capacity. These (( ))
To summarize, (( )) can be found in Tables 29-1, 29-2, 29-3a, 29-3b, 29-4, and 29-5 in WCAP-16996-P, Revision 1 (see pages 29-5 through 29-11). A compilation of the uncertainty parameter values and ranges can also be found in Table I of LTR-NRC-15-85. Also note that with either of these above References, (( )) as documented in LTR-NRC-15-102, Revision 2.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
Consistent with the NRC-approved methodology, the (( )) for the McGuire Units 1 and 2 and Catawba Unit 1 Region I analysis.
The NRC staff found that the licensee appropriately used the specified biased uncertainty parameters as noted above and, therefore, meets the requirement of Limitation and Condition 8.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation and Condition 9 - Effect of Bias in FSLOCATM EM Applications for Region I In Section 5.0, Table 22 of the safety evaluation states, in part, that:
In PWR plant type-specific applications of the FSLOCATM EM for designs which are not Westinghouse 3-loop PWRs, a confirmatory evaluation will be performed for Region I analyses to assess the effect associated with the (( )) This confirmatory evaluation will be performed once for each PWR plant type (e.g., Westinghouse design four-loop PWR plant) analyzed with the FSLOCATM EM and referenced in subsequent plant-specific FSLOCATM analyses of the same PWR plant type.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
McGuire Units 1 and 2 and Catawba Unit 1 are Westinghouse-designed 4-loop PWRs.
The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 5 [Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs) (Proprietary/Non-Proprietary), LTR-NRC-18-50, Revision 0, July 2018].
The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs),
dated July 2018 (ML18198A038 (package)). This document describes the sensitivity studies done on the selected parameters and demonstrates that (( )) Therefore, the NRC staff found that the licensee has met the requirements of Limitation and Condition 9.
Limitation and Condition 10 - Boundary Between FSLOCATM EM Region I and Region II Breaks In Section 5.0, Table 22 of its safety evaluation states, in part, that:
In PWR plant type-specific application of the FSLOCATM EM for designs which are not Westinghouse 3-loop PWRs, a confirmatory evaluation will be performed to demonstrate that the applied break size boundary between Region I and Region II serves the intended goal of
(( )) As of part this evaluation, it will be demonstrated that no unexplained behavior in the predicted safety criteria, including PCT, occurs across the boundary between Region I and Region II. In addition, it will be confirmed that (( )) In addition, it is important to also assure that the limiting small break between about 2-and 4-inch in an equivalent break diameter is properly captured by the robust Region I analysis approach. Plants with larger Reactor Coolant System (RCS) fluid volumes than the Beaver Valley plant test example in WCAP-16996-P/WCAP-16996-NP, Revision 1, should cover the same 2-to 4-inch range using break area to RCS volume scaling to assure that the 2-to 4-inch break range is preserved and not artificially truncated. This confirmatory evaluation will be performed once for each PWR plant type (e.g., Westinghouse design four-loop PWR plant) analyzed with the FSLOCATM EM and referenced in subsequent plant-specific FSLOCATM analyses of the same PWR plant type.
The WCOBRA/TRAC-TF2 code is applicable for analysis over the entire break spectrum of
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION LOCA transients. However, for the purpose of the Region II analysis, the minimum of the break area sampling should extend only to 1.0 ft2 consistent with the ASTRUM LBLOCA EM (WCAP-16009-P-A, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), Revision 0) in lieu of the Region I/II boundary.
Licensee Compliance Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
McGuire Units 1 and 2 and Catawba Unit 1 are Westinghouse-designed 4-loop PWRs.
The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 5 [Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs) (Proprietary/Non-Proprietary), LTR-NRC-18-50, Revision 0, July 2018.]. The minimum sampled break area for the McGuire Units 1 and 2 and Catawba Unit 1 Region II analysis is 1 ft2.
The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs).
This document describes the sensitivities performed to demonstrate that the boundary between Region I and Region II breaks is appropriate for a 4-loop Westinghouse-designed plant. In addition, the Region II analysis considers a minimum break area of 1.0 ft2 consistent with the requirement in the limitation and condition.
Therefore, the NRC staff found that the requirements of Limitation and Condition 10 are met as the necessary sensitivity studies to determine the appropriate break size range for Region I and boundary between Region I and Region II were performed.
Limitation and Condition 11 - (( Sample Size )) in FSLOCATM EM Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II In Section 5.0, Table 22 of the safety evaluation states, in part, that:
For each analysis performed using the FULL SPECTRUMTM LOCA methodology, the
(( )) seed, analysis inputs, and (( )) to be used for the Region I and Region II uncertainty analyses will be declared and documented prior to performing the uncertainty analyses.
The (( )) will not be adjusted as a result of the outcome. Should a plant-specific application of the FSLOCATM EM deviate from the originally declared analysis inputs for the intended purpose of demonstrating compliance with the applicable acceptance criteria, all modification(s) will be discussed in a calculation file and in the ECCS analysis submittal to NRC, as applicable, to explain the applicable reasons for the modification(s).
In this instance, the analysis inputs will be modified only for the purpose of reflecting the implemented and described modeling changes. In addition, the calculated preliminary values for PCT, MLO [ maximum local oxidation ], and CWO [ core wide oxidation ] for each such case will be summarized for information only in the ECCS analysis submittal to the NRC. Because these preliminary analyses and results are not intended to demonstrate compliance with the criteria of 10 CFR 50.46, formal Appendix B verification and archival documentation of the underlying analyses are not required.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Furthermore, operating ranges used in a plant-specific analysis as part of the sampling uncertainty analysis for Regions I and II are to be supplied for review by the NRC in a table format for both regions. In plant-specific reviews, the uncertainty treatment for such plant operating parameters including the sampled distributions and ranges will be considered acceptable if they meet or exceed corresponding design basis and/or Technical Specification limiting conditions for operation limits, with uncertainties included, as appropriate. Alternative approaches may be used, provided they are supported with appropriate justification. (( )) are given in Table 1 of LTR-NRC-17-47.
Note that (( )) as per limitation no. 15 below.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
This Limitation and Condition was met for the McGuire Units 1 and 2 and Catawba Unit 1 analysis as follows:
- 1.
The (( )) the Region I and Region II analysis seeds, and the analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty analyses. The (( )) and the Region I and Region II analysis seeds were not changed once they were declared and documented.
- 2.
The analysis inputs were not changed once they were declared and documented.
- 3.
The plant operating ranges which were sampled within the uncertainty analyses are provided for McGuire Units 1 and 2 and Catawba Unit 1 in Table 1 [of Enclosure 4 to the LAR].
The NRC reviewed the information related to initial conditions and plant operating ranges as presented in the LARs Enclosure 2 and found them appropriate for use in the methodology. In addition, given that the licensee has declared and documented the appropriate inputs and did not change these values once declared and documented, therefore; the NRC staff found that the licensee has met the requirements of Limitation and Condition 11.
Limitation and Condition 12 - Steam Generator Heat Removal During SBLOCAs In Section 5.0, Table 22 of the safety evaluation states, in part, that:
In plant-specific applications of the FSLOCATM EM, a check will be performed to confirm that effects associated with dynamic pressure losses from the steam generator secondary side to the main steam safety valves (MSSVs) are properly considered and adequately accounted for in the plant model used for the design-basis LOCA analyses consistent with NRC Information Notice 97-09, Inadequate Main Steam Safety Valve (MSSV) Set Points and Performance Issues Associated with Long MSSV Inlet Piping.
SBLOCA performance is dependent on secondary pressure as it establishes primary pressure, and the consequential emergency core cooling system injection rate and potential for and degree of core uncovery.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
A bounding plant-specific dynamic pressure loss from the [Steam Generator] SG secondary side to the MSSVs was modeled in the McGuire Units 1 and 2 and Catawba Unit 1 analysis.
As discussed in the response to request for additional information (RAI) 132 in Reference 17 [Submittal of Westinghouse Responses to WCAP-16996-P, 'Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)' Request for Additional Information - Set 8 RAIs 127, 132-135 (Proprietary), Project 700, TAC No. ME5244, LTR-NRC-14-4, Revision 0, January 2014.], (( )) To comply with this requirement, the initial opening pressure of the MSSV was modeled as the plant-specific second stage MSSV set pressure, plus uncertainty (1240.8 psia). For all three units, this value bounds the plant-specific first stage MSSV set pressure, plus uncertainty, plus the plant-specific dynamic pressure loss from the SG secondary side to the MSSVs during a SBLOCA transient.
The licensee computed a plant-specific dynamic pressure loss from the SG secondary side to the MSSVs during a SBLOCA transient. To account for this in the analyses, the licensee selected the MSSVs second stage set pressure (with +3% uncertainty) as the opening pressure for the MSSVs. Since this value is larger than the first stage opening set pressure (with
+3% uncertainty) and dynamic losses, the NRC staff found this acceptable and meets the requirements of Limitation and Condition 12.
Limitation and Condition 13 - Upper Head Spray Nozzle Loss Coefficient In Section 5.0, Table 22 states, in part, that:
In plant-specific applications of the FSLOCATM EM, (( )) in the PWR model used to perform the design-basis LOCA transient calculations, to capture the proper core two-phase level response should the core uncover. Additionally, the (( )) in such calculations. See Section 29.5.3, Venting, page 29-141 of WCAP-16996-P, Revision 1.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
The (( )) in the analysis for McGuire Units 1 and 2 and Catawba Unit 1. The (( )) in the analysis.
Based on its review of the above information, the NRC staff found that the licensee has met the requirements of Limitation and Condition 13.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation and Condition 14 - Correlation for Oxidation In Section 5.0, Table 22 of the safety evaluation states that:
For demonstration of compliance with the current 10 CFR 50.46 oxidation criterion, the oxidation result using Baker-Just to convert the LOCA transient time-at-temperature to an equivalent cladding reacted shall be compared against the 17 percent limit. If Cathcart-Pawel is used to convert the LOCA transient time-at-temperature to an equivalent cladding reacted, the oxidation result shall be compared to a 13 percent limit with the pre-transient oxide layer thickness being included in the prediction results.
Should this measure (Cathcart-Pawel) 13 percent limitation) not be carried forth to other NRC approvals of new realistic applications or should the value be changed, this SE and the two associated restrictions will be subsequently revised. See memorandum Ashok Thadani, Director, RES to Samual J. Collins, Director, NRR, Research Information Letter 0202, Revision of 10 CFR 50.46 and Appendix K, dated June 20, 2002, Appendix 2, page 9.
Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
In this analysis, the MLO acceptance criterion of 17% is replaced with the NRC-approved AXIOM cladding performance-based embrittlement acceptance criterion. The Cathcart-Pawel equivalent cladding reacted (ECR) is confirmed to remain below the DBT [ductile-to-brittle transition] limit for AXIOM cladding described in Section 3.11 of Reference 3 [ Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, WCAP-18546-P-A, Revision 0, March 2023 ]. Limitation and Condition Number 14 is therefore not applicable to the McGuire Units 1 and 2 and Catawba Unit 1 FSLOCA EM analysis.
Based on its review of the above information, NRC staff found that the FSLOCA EM Limitation and Condition 14 is not applicable when using AXIOM cladding.
Limitation and Condition 15 - LOOP [Loss of Offsite Power] versus OPA [Offsite Power Available] Treatment in FSLOCATM EM Uncertainty Analyses for Region II In Section 5.0, Table 22 of the safety evaluation states, in part, that:
Identification of the offsite power availability limiting condition for the Region II FSLOCATM evaluation is required by GDC 35. In lieu of the method proposed by Westinghouse for addressing this requirement described in LTR-NRC-15-102, Revision 2, page 25, plant-specific applications of the FSLOCATM EM should include two complete sets of sampled statistical evaluations; (1) a complete set with offsite power available and (2) a second complete sampling set without offsite power available.
For each set, the calculated statistical results at the 95/95 probability, confidence level should be demonstrated to comply with regulatory limits for PCT, MLO, and CWO. The
(( )) to provide the required 95/95 probability, confidence statement that addresses the three major criteria of PCT, MLO, and CWO. This condition should be consistent with limitation number 11 in the table for (( )) for each sample set.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Licensee Compliance In Section 2.3 of Enclosure 2 of its submittal dated August 18, 2025, the licensee states, in part, that:
The Region II uncertainty analysis for McGuire Units 1 and 2 and Catawba Unit 1 was performed twice; once assuming OPA and once assuming LOOP. The results from both analyses that were performed are in compliance with the applicable ECCS acceptance criteria (see Section 6.0). The (( ))
Given that the licensee has performed the Region II analysis for both LOOP and OPA using a sample size greater than the minimum required and that the NRC staff found those results acceptable for Limitation and Condition 15.
A regulatory finding summary of the NRC staffs review concerning the 15 Limitations and Conditions is made in Section 3.4 of this SE.
3.3.2 Composite model In Enclosure 4, Section 3.0, Composite Model Approach, the licensee stated that the current licensing basis LBLOCA analysis for McGuire Units 1 and 2 and CNS, Units 1 and 2, was performed with the Code Qualification Document (CQD) EM. This approach uses a composite model of the four units and was approved by NRC (ML003753895 and ML003756631, LAR Ref. 20 and 21, respectively). The licensee used this same approach for the FSLOCATM EM analysis for McGuire Units 1 and 2 and Catawba Unit 1.
During the regulatory audit, the NRC staff reviewed Westinghouse document WCAP-19088-P, Revision 0, Engineering Summary Report of the McGuire Units 1 and 2 and Catawba Unit 1 Loss-of-Coolant Accident (LOCA) Analysis with the FULL SPECTRUM LOCA (FSLOCA)
Methodology, September 2025. The NRC staff reviewed the summary of the composite model approach which describes the sensitivities and analysis used to determine the composite plant models details.
As part of this approach, the licensee developed two vessel models to capture the differences in the upper internals among the units. Previous sensitivity studies were performed and determined that the McGuire Unit 1 vessel model was limiting for the Code Qualification Document (CQD) LBLOCA analysis. The licensee confirmed that this model is also appropriate for use in both Region I and Region II FSLOCA analysis by performing additional sensitivity studies.
Previous sensitivity studies were also performed for accumulator parameters including line resistance and water volume. These sensitivities showed that minimum line resistance and minimum water volume were determined to be limiting for the CQD LBLOCA. The CQD LBLOCA also used the wider accumulator pressure range (as defined in TSs) for CNS, Units 1 and 2, which encompasses the range for MNS, Units 1 and 2. Similar sensitivity studies were performed for the Region II analysis with the FSLOCA EM, which showed similar trends as the sensitivity studies for the CQD LBLOCA analysis. Therefore, the licensee used the same accumulator modeling approach as used in the CQD LBLOCA analysis for the Region II analysis with the FSLOCA EM (minimum line resistance, minimum water volume range, wider
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION pressure range). Based on its review of the above information, the NRC staff finds the continued use of the CQD approach acceptable as it is consistent with use of the FSLOCA methodology.
For Region I, Figure 31.3-4 of the FSLOCA TR shows a correlation between accumulator pressure and PCT. Lower accumulator pressure delays the accumulator injection which delays the termination of the boil-off heatup for SBLOCA transients, leading to higher PCTs. The narrower accumulator pressure range for McGuire Units 1 and 2 results in lower sampled pressures and thus higher PCTs for Region I, so the narrower accumulator pressure range was used in Region I analysis. The licensee stated that accumulator line resistance and accumulator water volume have a negligible impact on the calculated results for Region I, so the same modeling approach used in the Region II analysis with the FSLOCA EM was also used in the Region I analysis with the FSLOCA EM (minimum line resistance, minimum water volume range).
Based on its review of the above information, the NRC staff finds that for the Region I analysis, the accumulators inject over a much longer period of time than in the Region II analysis and would not be expected to empty and therefore, the accumulator line resistance and water volume have a negligible impact on the results. Therefore, the NRC staff finds that the accumulator modelling approach for Region I is acceptable and consistent with the use of the FSLOCA methodology.
In Enclosure 1, Attachment 4 to the LAR, the licensee provided a comparison of the CQD inputs to those in the FSLOCATM EM. As stated by the licensee, several of the FSLOCA analysis inputs were changed from the CQD analysis to improve operating margins, account for instrument uncertainties, add conservatism to safety analysis margins, or maintain compliance with the new FSLOCA TM methodology. The NRC staff finds the changes from the existing CQD analysis acceptable for use as they are consistent with plant operating ranges and the FSLOCA EM.
The licensee stated that the ongoing process to assure the composite model remains representative and bounding for the units will continue to be followed. This process is defined in Duke Power letter to USNRC, License Amendment Request, Implementation of Best-Estimate Large Break Loss of Coolant Accident (BELBLOCA) Analysis Methodology, dated August 10, 2000 (ML003741728) and approved by the NRC (See ML003753895 and ML003756631).
Based on its review of the above information, NRC staff found that the composite model acceptable for use in the FSLOCA TM EM as the process gives adequate assurance that the composite plant analysis is qualitatively representative and quantitatively bounding for the three units (CNS, Unit 1, and MNS, Units 1 and 2) and that any significantly differing unit will be identified and analyzed on a plant-specific basis.
3.3.3 Results -vs-acceptance criteria To demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4), the following criteria must be met:
- 1. Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2,200°F.
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- 2. Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
- 3. Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
- 4. Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
The licensee presented the results for the FSLOCA TM analysis in Table 6 in the LARs and these are repeated in Table 1 below. For both PCT and CWO, the results from the analysis show the plant-specific values remain below the 10 CFR 50.46 acceptance criteria.
The maximum cladding oxidation acceptance criterion (17%) defined in 10 CFR 50.46 (b)(2) is replaced with the approved AXIOM cladding performance-based embrittlement acceptance criterion as stated in the AXIOM TR. The Cathcart-Pawel ECR is confirmed to remain below the ductile-to-brittle transition limit for AXIOM cladding described in Section 3.11 of the AXIOM topical report (WCAP-18546-P-A). This results in an acceptance criterion where the minimum ECR margin (MEM) is 0%.
Table 1. McGuire Units 1 and 2 and Catawba Unit 1 Analysis Results with the FSLOCA EM Parameter Region I Value Region II Value (OPA)
Region II Value (LOOP)
Acceptance Criteria PCT 1,199°F 1,662°F 1,647°F 2,200°F MEM 5.33%
5.07%
4.82%
0%
CWO 0.00%
0.09%
0.08%
1%
The coolable geometry acceptance criterion is assured by compliance with the first three acceptance criteria and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies. As discussed in Section 32.1 of the FSLOCA TM EM topical report (WCAP-16996-P-A) the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). The licensee stated that the FSLOCA TM EM analysis does not affect the existing calculations that support the analysis of record related to combined LOCA and seismic loads. The previous calculations on grid deformation due to combined LOCA and seismic loads remain valid and are described in Section 4.2.1.3.2, Fuel Assembly Structure, of the MNS Update Final Safety Analysis Report (UFSAR), version 25, (ML25083A313) and in Section 4.2.4.5, Fuel Assembly, of the CNS UFSAR, version 24, (ML24296A048, public) the UFSAR. Both of those UFSARs state that Grid crush analyses using combined seismic and LOCA loadings show that the fuel assembly will maintain a geometry that is capable of being cooled under the worst-case accident Condition IV event.
Therefore, NRC staff finds the coolable geometry criterion is met.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4 Technical Conclusion Based on its review of the proposed changes, the NRC staff concludes the following:
The addition of AXIOM as an acceptable fuel rod cladding material to TS 4.2.1 for CNS, Unit 1, and MNS, Units 1 and 2, is acceptable as the change is consistent with the relevant NRC-approved topical report.
The deletion of the note in TS 4.2.1 related to CNS lead assemblies is acceptable as the licensee is no longer considering using mixed oxide fuel.
The deletion of two legacy analytical methods, WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE (W Proprietary) and BAW-10168-P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants (B&W Proprietary), from TS 5.6.5 for CNS, Units 1 and 2, and MNS, Units 1 and 2, is acceptable as the methods are no longer used by the licensee as they were used with older fuel variants.
The deletion of the analytical method BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, in TS 5.6.5 for CNS is acceptable as the method is used with mixed oxide fuel that the licensee is no longer considering using.
The addition of a note to analytical methods WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, and WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis for CNS and MNS in TS 5.6.5 stating that the methods are not to be used to determine core operating limits once Optimized ZIRLO-clad fuel is fully discharged from the core is acceptable as these methodologies are being replaced with the FSLOCA EM.
The addition of NRC-approved analytical method WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) to TS 5.6.5 for CNS, Unit 1, and MNS, Units 1 and 2, is acceptable as NRC staff determined the licensee appropriately applied the FSLOCA TM EM, met all limitation and conditions, and the resulting analysis meets the acceptance criteria defined in 10 CFR 50.46 (b)(1) through (b)(4).
The addition of NRC-approved analytical method WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel to TS 5.6.5 for CNS, Unit 1, and MNS, Units 1 and 2, is acceptable given the licensee demonstrated compliance with the subject limitations and conditions. Further, this change also continues to meet 10 CFR 50.36(c)(4) and (c)(5) and include design features of the facility, which, if altered or modified, would have a significant effect on safety, and continues to provide TSs administrative controls.
Based on the above, the NRC concludes that the licensee would continue to meet the requirements of 10 CFR 50.36 and 10 CFR 50.46.
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4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the South Carolina and North Carolina State officials were notified of the proposed issuance of the amendment on February 19, 2026. The State of South Carolina officials had no comments on March 11, 2026 and the State of North Carolina had no comment on March 16, 2026.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 December 30, 2025 (90 FR 61169), 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.
Date: March 25, 2026
OFFICIAL USE ONLY - PROPRIETARY INFORMATION S. Gibby OFFICIAL USE ONLY - PROPRIETARY INFORMATION
SUBJECT:
CATAWBA NUCLEAR STATION, UNITS 1 AND 2, AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 323, 319; 333, 312, RESPECTIVELY, TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW USAGE OF A FULL SPECTRUM LOSS-OF-COOLANT-ACCIDENT (LOCA)
METHODOLOGY (EPID L-2025-LLA-0132) DATED MARCH 25, 2026 DISTRIBUTION:
PUBLIC RidsACRS_MailCTR Resource RidsNrrDorlLpl2-1 Resource RidsNrrLAKZeleznock Resource RidsNrrPMCatawba Resource RidsNrrPMMcGuire Resource RidsRgn2MailCenter Resource RidsNrrDssStsb Resource RidsNrrDssSnsb Resource RidsNrrDssSfnb Resource RBeaton, NRR CAshley, NRR KHeller, NRR ADAMS Accession Nos.:
ML26062A219 Package ML26062A213 Proprietary ML26062A215 Non-Proprietary 20260303-50016 e-concurrence Case NRR-058