RA-25-0014, License Amendment Request to Adopt the Full Spectrum Loss of Coolant Accident Methodology and Axiom Fuel Cladding Topical Reports and Associated 10 CFR 50.46 Exemption Request for Use of Axiom Fuel Cladding
| ML25230A072 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 08/18/2025 |
| From: | Gibby S Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| Shared Package | |
| ML25230A071 | List: |
| References | |
| RA-25-0014 | |
| Download: ML25230A072 (1) | |
Text
Shawn K. Gibby Vice President Nuclear Engineering Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 704-519-5138 Shawn.Gibby@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED August 18, 2025 10 CFR 50.12 Serial: RA-25-0014 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 / RENEWED LICENSE NOS. NPF-35 AND NPF-52 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17
Subject:
License Amendment Request to Adopt the Full Spectrum Loss of Coolant Accident Methodology and AXIOM Fuel Cladding Topical Reports and Associated 10 CFR 50.46 Exemption Request For Use of AXIOM Fuel Cladding Ladies and Gentlemen:
In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Unit 1 (CNS U1) and McGuire Nuclear Station, Units 1 and 2 (MNS).
The proposed amendment requests the addition of the Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-16996-P-A, Revision 1, Realistic LOCA [Loss of Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (ADAMS Accession No. ML17277A130), to the list of approved analytical methods used to determine the core operating limits provided in TS 5.6.5, Core Operating Limits Report (COLR). Additionally, this amendment proposes the annotation of select legacy LOCA methods and the deletion of others listed in TS 5.6.5.b to restrict their future use and allow for a staggered implementation during refueling outages at each unit.
The amendment also requests a modification to the TS to permit the use of the Westinghouse fuel cladding alloy designated as AXIOM. Specifically, the proposed amendment requests a revision to the TS 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 (ADAMS Accession No. ML23089A063) to the referenced analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding.
A separate license amendment request to address the adoption of the aforementioned Westinghouse topical reports for the FULL SPECTRUMTM LOCA (FSLOCATM) Evaluation
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-25-0014 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED Methodology (EM) and AXIOM fuel rod cladding for CNS Unit 2 (CNS U2) will be submitted at a later date. Inclusion of notes within the associated CNS TS are proposed to reflect applicability of the requested license amendment changes to only CNS U1. Deletion of legacy LOCA analysis methods is administrative and is applicable to both CNS U1 and CNS U2.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been concluded that the proposed changes involve no significant hazards consideration.
of this license amendment request provides Duke Energys evaluation of the proposed changes. Attachments 1 and 2 of the enclosure provide a copy of the existing TS pages and TS Bases pages, respectively, marked with the proposed changes. Attachments 3 and 4 of the enclosure provide additional information related to FSLOCATM plant input parameters. The TS Bases markups are provided for information only and will be incorporated in accordance with each respective sites TS Bases Control Program upon implementation of the approved license amendments. Enclosures 2 and 4 contain the Westinghouse proprietary and non-proprietary versions of a supporting document referenced in the evaluation. An affidavit from Westinghouse attesting to the proprietary nature of the information is provided in Enclosure
- 3. Duke Energy requests that Enclosure 2 be withheld from public disclosure in accordance with 10 CFR 2.390.
Furthermore, in accordance with the provisions of 10 CFR 50.12, Duke Energy is requesting an exemption from certain requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, as provided in Enclosure 5, to allow the use of the AXIOM fuel rod cladding for CNS U1 and MNS. The exemption request relates solely to the specific types of cladding material specified in the regulation for use in light water reactors. As written, the regulation presumes use of either Zircaloy or ZIRLO High Performance Fuel Cladding.
Approval of the proposed license amendment and corresponding exemption request is requested by March 31, 2026. Once approved, the amendment will be implemented for CNS U1 prior to the Unit 1 Cycle 30 reload campaign in Spring 2026 and implemented for MNS prior to the Unit 1 Cycle 32 reload campaign in Fall 2026. Revisions to the CNS and MNS Updated Final Safety Analysis Reports (UFSAR), necessary to reflect approval of this submittal, will be made in accordance with 10 CFR 50.71(e), with approved exemptions.
In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated North Carolina and South Carolina Officials.
This letter contains no regulatory commitments.
Please refer any questions regarding this submittal to Ryan Treadway, Director - Nuclear Fleet Licensing, at (980) 373-5873.
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-25-0014 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 2 THIS LETTER IS UNCONTROLLED I declare under penalty of perjury that the foregoing is true and correct.
Executed on August 18, 2025.
Sincerely, Shawn Gibby Vice President - Nuclear Engineering : Evaluation of the Proposed Amendment : Proposed Technical Specification Changes (Mark-up) : Proposed Technical Specification Bases Changes (Mark-up) : Plant Operating Parameters Compared to Technical Specification Limits : Comparison of CQD and FSLOCA Input Parameters : Application of Westinghouse FULL SPECTRUM LOCA Evaluation Model to McGuire Units 1 and 2 and Catawba Unit 1 (Proprietary Version) : Westinghouse Affidavit : Application of Westinghouse FULL SPECTRUM LOCA Evaluation Model to McGuire Units 1 and 2 and Catawba Unit 1 (Non-proprietary Version) : 10 CFR 50.46 Exemption Request for Use of AXIOM Fuel Cladding cc:
USNRC Region II - Regional Administrator USNRC Resident Inspector - CNS USNRC Senior Resident Inspector - MNS USNRC NRR Project Manager - Fleet USNRC NRR Project Manager - CNS USNRC NRR Project Manager - MNS L. Brayboy, Radioactive Materials Branch Manager - NC DHHS S. Jenkins, Director - Radiological Health Program - SC DES N. Gauthier, Manager - Nuclear Response Section - SC DES L. Garner, Manager - Radioactive & Infectious Waste Section - SC DES
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 ENCLOSURE 1 EVALUATION OF THE PROPOSED CHANGE 15 PAGES PLUS THE COVER AXIOM, FULL SPECTRUM, FSLOCA, ZIRLO, and Optimized ZIRLO are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world.
All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
U.S. Nuclear Regulatory Commission Page 1 of 15 Serial: RA-25-0014 Evaluation of the Proposed Change 1.0
SUMMARY
DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Unit 1 (CNS U1) and McGuire Nuclear Station, Units 1 and 2 (MNS).
The proposed amendment requests the addition of the Westinghouse Electric Company LLC (Westinghouse) Topical Report WCAP-16996-P-A, Revision 1, Realistic LOCA [Loss of Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (ADAMS Accession No. ML17277A130), to the list of approved analytical methods used to determine the core operating limits provided in TS 5.6.5, Core Operating Limits Report (COLR). Additionally, this amendment proposes the annotation of select legacy LOCA methods and the deletion of others listed in TS 5.6.5.b to restrict their future use and allow for a staggered implementation during refueling outages at each unit.
The amendment also requests a modification to the TS to permit the use of the Westinghouse fuel cladding alloy designated as AXIOM. Specifically, the proposed amendment requests a revision to the TS 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 (ADAMS Accession No. ML23089A063) to the referenced analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding. A corresponding exemption is being requested from the provisions of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, in order to support the use of this additional fuel rod cladding material. The exemption request for CNS U1 and MNS is provided in Enclosure 5 of this submittal.
A separate license amendment request to address the adoption of the aforementioned Westinghouse topical reports for the FULL SPECTRUM LOCA (FSLOCATM) Evaluation Methodology (EM) and AXIOM fuel rod cladding for CNS Unit 2 (CNS U2) will be submitted at a later date. Inclusion of Notes within the associated CNS TS are proposed to reflect applicability of the requested license amendment changes to only CNS U1. Deletion of legacy LOCA analysis methods is administrative and is applicable to both CNS U1 and CNS U2.
2.0 DETAILED DESCRIPTION
2.1 Background
System Design and Operation The primary function of the Emergency Core Cooling System (ECCS) following a LOCA is to remove the stored and fission product decay heat from the reactor core such that fuel rod damage, to the extent that it would impair effective cooling of the core, is prevented.
The principal mechanical components of the ECCS which provide core cooling immediately following a LOCA are the accumulators, the safety injection pumps, the centrifugal charging pumps, the residual heat removal pumps, refueling water storage tank, and the associated valves, and piping.
U.S. Nuclear Regulatory Commission Page 2 of 15 Serial: RA-25-0014 The ECCS is designed to cool the reactor core as well as to provide additional shutdown capability following initiation of the following accident conditions:
- 1. A pipe break or spurious valve lifting in the Reactor Coolant System (RCS) which causes a discharge larger than that which can be made up by the normal makeup system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS.
- 2. Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident.
- 3. Steam or feedwater system break accident including a pipe break or a spurious valve lifting in the steam system which would result in an uncontrolled steam release or a loss of feedwater.
- 4. A steam generator tube rupture.
AXIOM Fuel Cladding Westinghouse developed AXIOM fuel rod cladding material to provide enhanced corrosion resistance when compared to prior zirconium-based fuel cladding materials. The AXIOM alloy is the next generation of robust alloys targeting very high fuel duties. AXIOM cladding is designed to exhibit improved corrosion resistance, lower hydrogen pickup (HPU), and lower creep and growth compared to prior Westinghouse fuel cladding products, ZIRLO and Optimized ZIRLO.
AXIOM cladding is a niobium-bearing alloy with reduced tin content to increase corrosion resistance. It also includes vanadium and copper as alloying elements in order to improve specific properties such as HPU. 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.
2.2 Current TS Requirements A description of the current status of the associated TS is provided below.
2.2.1 CNS and MNS TS 4.2.1, Fuel Assemblies The current TS 4.2.1 for CNS and MNS states the following:
The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of ZIRLO, Optimized ZIRLO', 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.
The CNS TS is modified by a note that states: A maximum of four lead assemblies containing mixed oxide fuel and M5' cladding may be inserted into the Unit 1 or Unit 2 reactor core.
U.S. Nuclear Regulatory Commission Page 3 of 15 Serial: RA-25-0014 2.2.2 CNS and MNS TS 5.6.5, Core Operating Limits Report The existing TS 5.6.5 for MNS and CNS requires, in part, core operating limits be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and contains references to the approved analytical methods that are used to determine the core operating limits. The current methods listed in TS 5.6.5.b for LOCA analyses 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. This TS also references the associated topical report for Optimized ZIRLO, WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO.
2.3 Reason for the Proposed Change An analysis with the FSLOCA EM has been completed for MNS and CNS U1 to support 18-month cycle operation with AXIOM cladding. This analysis also supports an extended power uprate (EPU), though the EPU will not be implemented at this time. This license amendment request (LAR) for MNS and CNS U1 requests approval to apply the Westinghouse FSLOCA EM, which was developed to address the full spectrum of LOCAs which result from a postulated break in the RCS of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM 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 of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as small-break LOCAs (SBLOCAs) and large-break LOCAs (LBLOCAs).
By letter dated September 22, 2016 (ADAMS Accession No. ML16271A329), Duke Energy committed to submit LBLOCA analyses that apply NRC-approved methods including the effects of fuel pellet thermal conductivity degradation (TCD) to the NRC for review and approval. The Westinghouse FSLOCA EM includes the effects of TCD, and submittal of this LAR fulfills the Commitment for both MNS units and CNS U1.
AXIOM cladding is Westinghouses advanced cladding material. The AXIOM alloy is planned to be used as a fuel rod cladding material in all typical Westinghouse PWR production fuel assemblies. Only the cladding material is being changed; and it will be used with existing NRC-approved cladding dimensions, fuel structures, fuel assembly components, and fuel materials.
Duke is implementing AXIOM to gain cladding oxidation and hydrogen pickup margin. The margin gain will also assist Duke Energy with core design transitions for proposed future power uprates and 24-month fuel cycles at MNS and CNS U1.
2.4 Description of the Proposed Change A specific description of each change is provided below. In addition, a mark-up of the CNS and MNS TS for the following proposed changes is provided in Attachment 1 to this enclosure.
2.4.1 CNS and MNS TS 4.2.1, Fuel Assemblies This section is being revised to add the Westinghouse AXIOM alloy to the list of materials that may be used as the fuel rod cladding in CNS U1 and MNS fuel assemblies. The proposed
U.S. Nuclear Regulatory Commission Page 4 of 15 Serial: RA-25-0014 revised specification (with changes in bold and strikethroughs) for MNS TS 4.2.1 reads as follows:
The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either Zircalloy, ZIRLO, or Optimized ZIRLO', 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 ZIRLO', 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.
CNS TS 4.2.1 will receive the same proposed revisions to reflect the addition of the AXIOM alloy to the list of materials, with an additional clarifier that it is only applicable for Unit 1. There is no change proposed to the CNS-specific note addressing mixed oxide fuel and M5 cladding.
2.4.2 CNS and MNS TS 5.6.5, Core Operating Limits Report The proposed change revises MNS TS 5.6.5 to reflect the addition of Westinghouse topical reports WCAP-16996-P-A and WCAP-18546-P-A to the list of approved methods as follows:
- 20.
WCAP-16996-P-A, Revision 1, 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)
Since this proposed amendment will also apply to CNS U1, and not CNS U2, the proposed change for CNS TS 5.6.5 similarly reflects the addition of Westinghouse topical reports WCAP-16996-P-A and WCAP-18546-P-A to the list of approved methods, but also includes annotation reflecting that these topical reports are only applicable for Unit 1. This note will remain until a license amendment is issued for CNS U2 to similarly adopt these two topical reports.
- 21.
WCAP-16996-P-A, Revision 1, 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, Revision 0, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, (Westinghouse Proprietary) [For use by Unit 1 only.]
This proposed amendment also addresses administrative changes to clean up the list of analytical methods provided in CNS and MNS TS 5.6.5. For both CNS and MNS, two legacy LOCA methodologies are proposed for deletion, since they 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 LBLOCA analysis in 2000 which replaced the BASH LBLOCA method (WCAP-10266-P-A).
U.S. Nuclear Regulatory Commission Page 5 of 15 Serial: RA-25-0014
- 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)
For Catawba, it is proposed to also delete the fuel rod design methodology below, since it only supported Framatome Mixed Oxide (MOX) fuel Lead Test Assemblies, which are no longer under consideration for use as batch feed fuel.
- 17.
BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, (Framatome ANP Proprietary).
In addition to the proposed deletion of these legacy methodologies, it is also proposed to annotate the following two LOCA methodologies to indicate when they will no longer be used to establish core operating limits once all of the Optimized ZIRLO cladding is discharged from the core designs. The proposed schedule for implementation of AXIOM cladding fuel is shown below for each affected unit. Assuming normal reload core design practices, the Optimized ZIRLO cladding fuel would be fully discharged from the core designs two cycles after the implementation of AXIOM cladding.
Unit Proposed AXIOM Implementation Full Cores of AXIOM Catawba Unit 1 Spring 2026, Cycle 30 Spring 2029, Cycle 32 McGuire Unit 1 Fall 2026, Cycle 32 Fall 2029, Cycle 34 McGuire Unit 2 Fall 2027, Cycle 32 Fall 2030, Cycle 34 CNS:
- 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 after startup of Catawba Unit 1 Cycle 32].
- 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 after startup of Catawba Unit 1 Cycle 32].
MNS:
- 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 after startup of McGuire Unit 1 Cycle 34 and McGuire Unit 2 Cycle 34].
- 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
U.S. Nuclear Regulatory Commission Page 6 of 15 Serial: RA-25-0014 to determine core operating limits after startup of McGuire Unit 1 Cycle 34 and McGuire Unit 2 Cycle 34]
3.0 TECHNICAL EVALUATION
FSLOCA The technical evaluation for the application of the Westinghouse FSLOCA EM to MNS and CNS U1 is provided in Enclosures 2 (Proprietary) and 4 (Non-proprietary). The application of the FSLOCA EM to MNS and CNS U1 is consistent with the NRC-approved methodology provided in WCAP-16996-P-A, as modified for AXIOM cladding in WCAP-18546-P-A. The application of this topical report to MNS and CNS U1 involves the performance of a composite FSLOCA EM analysis that covers both MNS units and CNS U1.
The proposed amendment updates the listing of approved analytical methods in TS 5.6.5.b.
These changes are administrative in nature because the updated list of analytical methods will continue to ensure core operating limits can be established. The updated listing will reflect the adoption of WCAP-16996-P-A, Revision 1, demonstrating the compliance with the ECCS performance criterion of 10 CFR 50.46, subject to the NRC-specified limitations and conditions in the topical report SE. Additionally, the updated listing will reflect the removal of legacy methods that no longer support current fuel designs, and the addition of an annotation to the existing LOCA evaluation methods that continue to support Optimized ZIRLOTM cladding to restrict future use once all Optimized ZIRLO-clad fuel is discharged from the core designs.
These changes have no technical impact on the ability to meet COLR limits.
Of note, Duke Energy has an existing Commitment to re-analyze LBLOCA with methods that explicitly consider fuel pellet thermal conductivity degradation (TCD). The Commitment is stated in Duke Energy letter dated September 22, 2016 (ADAMS Accession No. ML16271A329), and is copied below.
Duke Energy will submit to the NRC for review and approval a LBLOCA analysis for Catawba and McGuire that apply NRC-approved methods that include the effects of fuel thermal conductivity degradation (TCD). The revised due date for this commitment is 24 months from the latest NRC approval of WCAP-17642-P, WCAP-16996-P, and any supplements that are needed for analysis related to the new 10 CFR 50.46(c) rule.
The FSLOCA methodology explicitly considers TCD effects. Submission of the subject license amendment request will satisfy the Commitment for MNS and CNS U1. A future CNS U2 license amendment request will similarly request the adoption of the FSLOCA EM and will satisfy the Commitment for the station.
AXIOM The NRC Staff reviewed Westinghouse topical report for AXIOM fuel rod cladding, WCAP-18546-P-A, Revision 1, and concluded in the issued safety evaluation (SE) dated December 16, 2022 (ADAMS Accession Nos. ML22320A685 (Non-Proprietary) and ML22306A275 (Proprietary)) that the generic topical report was acceptable for licensing applications, subject to the ranges of fuel types, cladding, and reactors identified in the SE being addressed by the licensees. Duke Energys proposed implementation of AXIOM fuel rod cladding complies with
U.S. Nuclear Regulatory Commission Page 7 of 15 Serial: RA-25-0014 the NRC limitations and conditions specified in Section 4.0 of the SE for the AXIOM topical report as follows:
Limitation and Condition 1 It is specified that AXIOM cladding must be used with the NRC-approved PWR designs. Both MNS units are PWRs that are currently licensed for operation by the NRC per operating licenses NPF-9 and NPF-17, respectively. Similarly, CNS U1 is a PWR that is currently licensed for operation by the NRC per operating license NPF-35. Duke Energys use of AXIOM cladding will not challenge this limitation related to PWR design.
Limitation and Condition 2 It is specified that AXIOM cladding must be used with the NRC-approved Westinghouse and CE fuel designs with corresponding pellet and assembly dimensions. Both MNS and CNS U1 utilize the 17 x 17 Westinghouse RFA design, an NRC-approved fuel design. Duke Energys use of AXIOM cladding will not challenge this limitation related to fuel design.
Limitation and Condition 3 It is specified that AXIOM cladding must be used with the NRC-approved fuel materials and pellet coatings or additives (e.g., ADOPT IFBA, gadolinium). 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.
Limitation and Condition 4 A fuel limitation is specified for AXIOM use which requires that fuel burnup shall currently be limited to 62 GWd/MTU peak rod average for all cladding types. 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 U1 will apply a peak rod average burnup limit of 62 GWd/MTU. This limitation will be validated by cycle-specific engineering analyses.
Limitation and Condition 5 A fuel limitation is specified for AXIOM use that requires the Best Estimate Oxide Thickness remain below 100 m. As provided in Section 5.1.1 of WCAP-18546-P-A, 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.
Limitation and Condition 6 A limitation is specified for the best estimate Hydrogen Pickup (HPU) over the operating life of the fuel. As shown in Section 5.2 of WCAP-18546-P-A, the overall maximum hydrogen content for AXIOM is significantly less than the value specified in Limitation and Condition 6, which demonstrates AXIOMs low HPU. This limitation will be validated by cycle-specific engineering analyses to ensure compliance with this condition and limitation.
Furthermore, the content below describes the impact of AXIOM cladding on other safety analyses and methods owned by Duke Energy, similar to the discussion in Sections 3.8.5 and 3.8.6 of the SE for WCAP-18546-P-A.
U.S. Nuclear Regulatory Commission Page 8 of 15 Serial: RA-25-0014 Containment Integrity Analyses The short-term LOCA mass and energy releases (M&E) are used to determine the maximum differential pressure for structural analyses within sub-compartments inside the containment building resulting from postulated pipe ruptures in the primary system piping. This transient lasts for 1 to 3 seconds in duration and the cladding material does not influence the mass and energy releases. The short-term LOCA M&Es are performed with Westinghouse methods, which are evaluated as being unaffected by the use of AXIOM cladding per Section 6.2.3.1 of WCAP-18546-P-A and Section 3.8.5.1 of the corresponding SE.
For long-term LOCA M&E release calculations, Duke Energy has licensed the methodology described in DPC-NE-3004-P-A, McGuire and Catawba Nuclear Stations Mass and Energy Release and Containment Response Methodology, used for containment integrity, maximum sump temperature, and equipment qualification at CNS and MNS. It is conservative for the long term LOCA M&E to maximize the rate of transfer of energy from the core into the coolant and out of the break. There is no hot rod or hot assembly modeled when generating long term LOCA M&E, and therefore fuel pellet and cladding interaction and rod burst are not modeled. Fuel thermal performance characteristics are adjusted to maximize fuel temperature and core stored energy. The AXIOM thermal material properties, as provided in Section 3.2 of WCAP-18546-P-A, are similar to other zirconium-based cladding materials, such that the long-term LOCA mass and energy releases are not affected by the use of AXIOM cladding.
The short-term steam line break (SLB) M&Es are performed with Westinghouse methods, which are evaluated as being unaffected by the use of AXIOM cladding per Section 6.2.3.3 of WCAP-18546-P-A, and Section 3.8.5.2 of the corresponding SE.
For long-term SLB M&E release calculations, Duke Energy has licensed the methodology described in DPC-NE-3004-P-A, used for containment integrity and equipment qualification at CNS and MNS. It is conservative for the long term SLB M&E to maximize the rate of transfer of energy from the core into the primary coolant, and maximize primary-to-secondary heat transfer for steam releases out of the secondary-side break. Fuel thermal performance characteristics are adjusted to maximize fuel temperature and core stored energy. The AXIOM thermal material properties, as provided in Section 3.2 of WCAP-18546-P-A, are similar to other zirconium-based cladding materials, such that the long-term SLB mass and energy releases are not affected by the use of AXIOM cladding.
Radiological Consequences Analyses Implementation of AXIOM fuel rod cladding will have no impact on models and method used in performing offsite and control room radiological dose consequences analyses for accidents.
Radiological consequence analysis does not model cladding. Change of cladding material could impact input to accident radiological consequences. Radiological consequences analyses consider the extent of fuel cladding damage resulting from postulated accidents. The analysis would be incorporated in a plant specific analysis using methods consistent with the analysis of record.
U.S. Nuclear Regulatory Commission Page 9 of 15 Serial: RA-25-0014
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements and Guidance 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors 10 CFR 50.46 requires, in part, that the calculated ECCS performance for light-water nuclear power reactors with zircaloy or ZIRLO fuel cladding meet certain criteria set forth in 10 CFR 50.46(b). Enclosure 5 of this submittal contains an exemption request being submitted in accordance with 10 CFR 50.12, requesting exemption from certain requirements of 10 CFR 50.46 to allow use of the AXIOM alloy for fuel rod cladding.
As it relates to the proposed change to adopt the FSLOCA EM in WCAP-16996-P-A, the FSLOCA EM satisfies the requirements of 10 CFR 50.46(b) paragraphs (1) through (4), which 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.
The proposed changes either continue to meet the requirements of this regulation or an exemption is justified as described in Enclosure 5.
10 CFR 50.36, Technical specifications The NRC's regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TS include items in the following five specific categories: (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.
The regulation does not specify the particular requirements to be included in a plants TS.
10 CFR 50.36(c)(4) states in part, that 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 other categories of the regulation. The proposed changes continue to meet the requirements of this regulation.
10 CFR 50.36(c)(5) states, 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. Each licensee shall submit any reports to the
U.S. Nuclear Regulatory Commission Page 10 of 15 Serial: RA-25-0014 Commission pursuant to approved technical specifications as specified in § 50.4. The proposed changes continue to meet the requirements of this regulation.
10 CFR Part 50, Appendix A, General Design Criteria (GDC) 10, 15, and 35 10 CFR 50, Appendix A, GDC 10 (Reactor design) states 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.
10 CFR 50, Appendix A, GDC 15 (Reactor coolant system design) states 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.
10 CFR 50, Appendix A, GDC 35 (Emergency core cooling) states 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. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure. Conformance with GDC 35 is described in more detail in the FSLOCA topical report, WCAP-16996-P-A, Revision 1.
The proposed changes do not affect compliance with these regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.
Conclusion The proposed change ensures plant compliance with the above regulations and guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.
4.2 Precedents AXIOM The proposed amendment requests a revision to the TS to update the description of fuel assemblies specified in TS 4.2.1, Fuel Assemblies, and add Westinghouse Topical Report WCAP-18546-P-A to the referenced analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding. A corresponding exemption is being requested from the provisions of 10 CFR 50.46 in order to support the use of this additional fuel rod cladding material. The NRC previously issued a license amendment to Turkey Point Nuclear Generating, Unit Nos. 3 and 4 (Turkey Point), by letter dated February 12, 2025 (ADAMS Accession No. ML25043A428) that
U.S. Nuclear Regulatory Commission Page 11 of 15 Serial: RA-25-0014 addressed revisions to the licensing basis to incorporate AXIOM cladding. The Turkey Point submittal also addressed several other changes that are not included within the scope of this request, including incorporation of additional advanced fuel features for pellets and fuel skeleton, application of an updated instrument channel uncertainty evaluation methodology, and changes to facilitate the transition to 24-month fuel cycles.
FSLOCA The proposed amendment requests the addition of Westinghouse Topical Report WCAP-16996-P-A, Revision 1, to the list of approved analytical methods used to determine the core operating limits provided in TS 5.6.5 for MNS and CNS U1. The NRC has previously issued license amendments for the following sites that have similarly requested approval to include this topical report within their respective TS COLR reference lists:
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak),
Amendment Nos. 185 and 185 to Renewed Facility Operating License Nos. NPF-87 and NPF-89, respectively, per letter dated December 20, 2023 (ADAMS Accession No. ML23319A387)
Beaver Valley Power Station, Unit Nos. 1 and 2 (Beaver Valley), Amendment Nos. 322 and 212 to Renewed Facility Operating License Nos. DPR-66 and NPF-73, respectively, per letter dated October 2, 2023 (ADAMS Accession No. ML23198A359). The Beaver Valley license amendment also added a note to the LOCA methods listed in the TS COLR list of references to restrict their future use.
Turkey Point Nuclear Generating Unit Nos. 3 and 4, Amendment Nos. 296 and 289 to Renewed Facility Operating License Nos. DPR-31 and DPR-41, respectively, per letter dated May 24, 2022 (ADAMS Accession No. ML22028A066).
The Comanche Peak and Beaver Valley license amendments also addressed the removal of Zircalloy from the list of fuel rod cladding in TS 4.2.1, Fuel Assemblies, which is beyond the scope of this request for MNS and CNS U1. Additionally, the Comanche Peak license amendment addressed a change to TS Safety Limit (SL) 2.1.1.2 in Reactor Core SLs to reflect the peak fuel centerline melt temperature specified in the Westinghouse performance analysis and design model (PAD5). This particular change is under NRC review for MNS and CNS in a separate submittal (ADAMS Accession No. ML25070A183) and is outside the scope of this license amendment request.
In reviewing the associated submittals for the precedents above, it was determined that additional information was needed as it (1) related to how FSLOCA plant input parameters correspond to TS Limiting Condition for Operation values, and (2) how FSLOCA plant input parameters may differ from previous inputs used in existing LBLOCA analyses. This information for MNS and CNS U1 can be found in Attachments 3 and 4 of this enclosure.
4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Unit 1 (CNS U1) and McGuire Nuclear Station, Units 1 and 2 (MNS). The proposed amendment requests the addition of the Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-
U.S. Nuclear Regulatory Commission Page 12 of 15 Serial: RA-25-0014 16996-P-A, Revision 1, Realistic LOCA [Loss of Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology)
(ADAMS Accession No. ML17277A130), to the list of approved analytical methods used to determine the core operating limits provided in TS 5.6.5, Core Operating Limits Report (COLR). Additionally, this amendment proposes the annotation of select legacy LOCA methods and the deletion of others listed in TS 5.6.5.b to restrict their future use and allow for a staggered implementation during refueling outages at each unit.
The amendment also requests a modification to the TS to permit the use of the Westinghouse fuel cladding alloy designated as AXIOM. Specifically, the proposed amendment requests a revision to the TS 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 (ADAMS Accession No. ML23089A063) to the referenced analytical methods in TS 5.6.5.b to allow the use of AXIOM alloy for fuel rod cladding. A corresponding exemption is being requested from the provisions of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, in order to support the use of this additional fuel rod cladding material. The exemption request for CNS U1 and MNS is provided in Enclosure 5 of this submittal.
A separate license amendment request to address the adoption of the aforementioned Westinghouse Topical Reports for the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology (EM) and AXIOM fuel rod cladding for CNS Unit 2 (CNS U2) will be submitted at a later date. Inclusion of notes within the associated CNS TS are proposed to reflect applicability of the requested license amendment changes to only CNS U1. Deletion of legacy LOCA analysis methods is administrative and is applicable to both CNS U1 and CNS U2.
Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
(1)
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves a change to utilize the FSLOCA EM, along with the implementation of associated TS changes. The proposed change to TS 5.6.5 permits the use of an NRC-approved methodology for analysis of the LOCAs to determine whether MNS and CNS U1 continue to meet the applicable design and safety analysis acceptance criteria. Restrictions on the future use of legacy LOCA methods are listed in TS 5.6.5.b in order to allow for a staggered implementation during refueling outages at each unit and has no direct impact upon plant operation or configuration. The results of the composite LOCA analyses for MNS and CNS U1 demonstrate the continued fulfilment of the 10 CFR 50.46(b)(1-4) emergency core cooling system (ECCS) performance acceptance criteria using an NRC-approved EM. These changes do not alter plant equipment nor the manner in which equipment is operated and maintained.
The proposed amendment also involves a change to TS 4.2.1 to allow the use of AXIOM-clad nuclear fuel by MNS and CNS U1, and a change to TS 5.6.5 to add the
U.S. Nuclear Regulatory Commission Page 13 of 15 Serial: RA-25-0014 NRC-approved topical report WCAP-18546-P-A for AXIOM fuel rod cladding to the COLR references. This topical report demonstrates that AXIOM fuel rod cladding has essentially the same properties as Optimized ZIRLOTM fuel rod cladding. Use of AXIOM fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The fuel cladding is not an accident initiator and does not affect accident probability. Use of AXIOM cladding meets the fuel design acceptance criteria and therefore does not significantly affect the consequences of an accident.
The proposed changes do not impact either the initiation of an accident or the mitigation of its consequences. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure. As a result, the outcomes of accidents previously evaluated are unaffected.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment involves a change to utilize the FSLOCA EM, along with the implementation of associated TS changes and restrictions on the future use of legacy LOCA methods. These changes will not create the possibility of a new or different accident due to credible new failure mechanisms, malfunctions, or accident initiators not previously considered. No physical plant modifications are being made as a result of the change to utilize the FSLOCA EM.
The proposed amendment also involves a change to TS to allow the use of AXIOM-clad nuclear fuel by MNS and CNS U1 and the addition of the NRC-approved topical report for AXIOM fuel rod cladding to the COLR references. This topical report demonstrates that AXIOM fuel rod cladding has essentially the same properties as Optimized ZIRLO fuel rod cladding. In performing similarly to the current fuel rod cladding, this precludes the possibility of the fuel rod cladding becoming an accident initiator and causing a new or different kind of accident.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated in the Updated Final Safety Analysis Report. No new single failure mechanisms will be created as a result of the proposed changes, and there are no alterations to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
U.S. Nuclear Regulatory Commission Page 14 of 15 Serial: RA-25-0014 (3)
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment involves a change to utilize the FSLOCA EM, along with the implementation of associated TS changes and restrictions on the future use of legacy LOCA methods. The analytic technique to be used in the analysis realistically describes the expected behavior of the MNS and CNS U1 reactor system during a postulated LOCA. Uncertainties have been accounted for as required by 10 CFR 50.46, and analysis shows that there is a high level of probability that all criteria contained in 10 CFR 50.46(b) are met. Approved methodologies would continue to be used to ensure that MNS and CNS U1 continue to meet applicable design criteria and safety analysis acceptance criteria.
The proposed amendment also involves a change to TS to allow the use of AXIOM-clad nuclear fuel by MNS and CNS U1 and the addition of the NRC-approved topical report for AXIOM fuel rod cladding to the COLR references. This topical report demonstrates that the material properties of the AXIOM fuel rod cladding are similar to the currently utilized Optimized ZIRLO fuel rod cladding. AXIOM is expected to perform similarly to Optimized ZIRLO for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. The use of AXIOM fuel rod cladding will not result in adverse changes to the operation or configuration of the facility. The proposed changes do not affect the acceptance criteria for any UFSAR safety analysis analyzed accidents or anticipated operational occurrences. All required safety limits will continue to be analyzed using methodologies approved by the NRC. This ensures that applicable design and performance criteria associated with the safety analysis will continue to be met and that the margin of safety is not affected.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Based upon the above evaluation, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVIRONMENTAL CONSIDERATION
S Duke Energy has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, Standards for protection against radiation, or it would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant
U.S. Nuclear Regulatory Commission Page 15 of 15 Serial: RA-25-0014 hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 1 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) 6 PAGES PLUS THE COVER
Design Features 4.0 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 eitheU=ircalloy, 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.
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.
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:
(continued)
Catawba Units 1 and 2 4.0-1 Amendment Nos. 284/280
Reporting Requirements 5.6 (continued)
Catawba Units 1 and 2 5.6-3 Amendment Nos. 222/217 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.
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).
Deleted.
Deleted.
Reporting Requirements 5.6 (continued)
Catawba Units 1 and 2 5.6-4 Amendment Nos 284/280.
5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
DPC-NE-2008-P-A, "Fuel Mechanical Reload Analysis
Methodology Using TACO3" (DPC Proprietary).
WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation
Model Using the NOTRUMP Code (W Proprietary).
DPC-NE-2009-P-A, Westinghouse Fuel Transition Report (DPC
Proprietary).
WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code
Qualification Document for Best-Estimate Loss of Coolant
Analysis (W Proprietary).
DPC-NE-1005P-A, Duke Power Nuclear Design Methodology
Using CASMO-4/SIMULATE-3 MOX, (DPC Proprietary).
BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code,(Framatome ANP Proprietary).
DPC-NE-1007-PA, Conditional Exemption of the EOC MTC
Measurement Methodology (Duke and Westinghouse
Proprietary).
WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core
Report, April 1995 (Westinghouse Proprietary).
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO', July 2006 (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.
Deleted.
INSERT 21 and 22 from below.
[Shall not be used to determine core operating limits after startup of Catawba Unit 1 Cycle 32.]
- 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.]
Design Features 4.0 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, 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.
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; McGuire Units 1 and 2 4.0-1 Amendment Nos. 288/267
Reporting Requirements 5.6 (continued)
McGuire Units 1 and 2 5.6-3 Amendment Nos. 226 / 208 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.
WCAP-10266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE, (W Proprietary).
- 3.
BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B& W Proprietary).
- 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)
Deleted.
Deleted.
Add: [Shall not be used to determine core operating limits after startup of McGuire Unit 1 Cycle 34 and McGuire Unit 2 Cycle 34.]
Reporting Requirements 5.6 (continued)
McGuire Units 1 and 2 5.6-4 Amendment Nos. 313/292 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).
- 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).
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 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.
Add: [Shall not be used to determine core operating limits after startup of McGuire Unit 1 Cycle 34 and McGuire Unit 2 Cycle 34.]
INSERT 20 and 21 from below.
- 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).
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 2 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP) 80 PAGES PLUS THE COVER (FOR INFORMATION ONLY)
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ESFAS Instrumentation B 3.3.2 BASES Catawba Units 1 and 2 B 3.3.2-6 Revision No. 14 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
- 1.
Safety Injection Safety Injection (SI) provides two primary functions:
- 1.
Primary side water addition to ensure maintenance or recovery of reactor vessel water level (coverage of the active fuel for heat removal, clad integrity, and for limiting peak clad temperature to < 2200qF); and
- 2.
Boration to ensure recovery and maintenance of SDM (keff < 1.0).
These functions are necessary to mitigate the effects of high energy line breaks (HELBs) both inside and outside of containment.
The SI signal is also used to initiate other Functions such as:
x Phase A Isolation; x
Containment Purge and Exhaust Isolation; x
Reactor Trip; x
Turbine Trip; x
Feedwater Isolation; x
Start of motor driven auxiliary feedwater (AFW) pumps; x
Start of control room area ventilation filtration trains; x
Enabling automatic switchover of Emergency Core Cooling Systems (ECCS) suction to containment sump; x
Start of annulus ventilation system filtration trains; x
Start of auxiliary building filtered ventilation exhaust system trains; x
Start of diesel generators meeting the 10 CFR 50.46 acceptance criteria (Ref. 16 and Ref. 17)
ESFAS Instrumentation B 3.3.2 BASES Catawba Units 1 and 2 B 3.3.2-52 Revision No. 14 REFERENCES (continued)
- 15. to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing.
- 16. 10 CFR 50.46.
- 17. WCAP-18546-P-A, March 2023.
Catawba Units 1 and 2 B 3.5.1-1 Revision No. 4 Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5.1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.
The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.
In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water.
The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.
Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series. The motor operated isolation valves are interlocked by P-11 with the pressurizer pressure measurement channels to ensure that the valves will automatically open as RCS pressure increases to above the permissive circuit P-11 setpoint.
This interlock also prevents inadvertent closure of the valves during normal operation prior to an accident. The valves will automatically open, however, as a result of an SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be open large break Initial accumulator inventory which is injected into the reactor vessel is lost out the break.
large break
Accumulators B 3.5.1 BASES Catawba Units 1 and 2 B 3.5.1-2 Revision No. 4 BACKGROUND (continued) and power removed during unit operation. In that the subject valves are normally open and do not serve as an active device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE)
Standard 279-1971 (Ref. 1) is not applicable in this situation. Therefore, the subject valve control circuit is not designed to this standard.
The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.
APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.
In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.
The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.
As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.
large break (for loss of offsite power assumption) for a large break LOCA in the modeling The largest break area considered for a large break LOCA is a double ended guillotine break in the RCS cold leg.
Accumulators B 3.5.1 BASES Catawba Units 1 and 2 B 3.5.1-3 Revision No. 4 APPLICABLE SAFETY ANALYSES (continued)
The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase.
This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a small break LOCA and there is a high level of probability that the criteria are met following a large break LOCA:
- a.
Maximum fuel element cladding temperature is d 2200qF;
- b.
Maximum cladding oxidation is d 0.17 times the total cladding thickness before oxidation;
- c.
Maximum hydrogen generation from a zirconium water reaction is d 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; and
- d.
Core is maintained in a coolable geometry.
Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute directly to the long term cooling requirements of 10 CFR 50.46. However, the boron content of the accumulator water helps to maintain the reactor core subcritical after reflood, thereby eliminating fission heat as an energy source for which cooling must be provided.
For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. The large and small break LOCA analyses are performed with accumulator volumes that are consistent with the LOCA evaluation models. To allow for operating margin, values of +/- 30 ft3 are specified.
The minimum boron concentration setpoint is used in the post LOCA subcriticality verification during the injection phase. For each reload is assumed to inject into the RCS intermediate acceptance criteria (Ref. 3 and Ref. 8) are met.
large break LOCA and the recovery phase of a small break LOCA For Unit 1, the large and small break LOCA analyses use a range of accumulator water volumes of 6790 gallons to 7422 gallons per the approved method (Ref.
9). For Unit 2, the large break LOCA analysis uses a range of accumulator water volumes of 7550 gallons to 8159 gallons, and the small break LOCA analysis uses a nominal accumulator water volume of 7855 gallons, per approved methods.
Both large and small break LOCA analyses use a nominal accumulator line volume from the accumulator to the check valve.
Accumulators B 3.5.1 BASES Catawba Units 1 and 2 B 3.5.1-4 Revision No. 4 APPLICABLE SAFETY ANALYSES (continued) cycle, the all rods out (ARO) critical boron concentration is verified to be less than the minimum allowed cold leg accumulator boron concentration.
The minimum boron concentration setpoint is also used in the post LOCA sump boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment with all rods in, minus the highest worth rod out (ARI N-1). Of particular interest is the large cold leg break LOCA, since boron accumulation in the core will be maximized during the cold leg recirculation phase due to core boiling.
The accumulation of boron in the core prevents the boron from returning to the sump, which leads to a boron diluted sump condition which may cause the core to become re-critical when switching over to hot leg recirculation. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. In particular, the equilibrium sump pH should be at least 7.5 following the design basis LOCA.
The large and small break LOCA analyses are performed with accumulator pressures that are consistent with the LOCA evaluation models. To allow for operating margin and accumulator design limits, a range from 585 psig to 678 psig is specified. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity.
The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Ref. 4).
The accumulators satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).
LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 3) could be violated.
For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and For Unit 1, the large break LOCA analysis uses a range of accumulator nitrogen cover pressures of 555 psig to 708 psig, and the small break LOCA analysis uses a range of accumulator nitrogen cover pressures of 555 psig to 669 psig, per the approved method (Ref. 9). For Unit 2, the large break LOCA analysis uses a range of accumulator nitrogen cover pressures of 555 psig to 708 psig, and the small break LOCA analysis uses a minimum accumulator nitrogen cover pressure of 555 psig, per approved methods.
large break Ref. 3 and Ref. 8
Accumulators B 3.5.1 BASES Catawba Units 1 and 2 B 3.5.1-5 Revision No. 4 LCO (continued) nitrogen cover pressure must be met. Additionally, the nitrogen and liquid volumes between accumulators must be physically separate.
APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.
This LCO is only applicable at pressures > 1000 psig. At pressures d 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 3) limit of 2200qF for small break LOCAs and there is a high level of probability that the peak cladding temperature does not exceed 2200qF for large break LOCAs.
In MODE 3, with RCS pressure d 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are allowed to be closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.
ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood.
Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break for the plant. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron concentration to within limits.
the 10 CFR 50.46 acceptance criteria (Ref. 3 and Ref. 8) are met
Accumulators B 3.5.1 BASES Catawba Units 1 and 2 B 3.5.1-8 Revision No. 4 SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification that power is removed from each accumulator isolation valve operators for NI54A, NI65B, NI76A, and NI88B when the RCS pressure is
> 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is d 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.
Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.
REFERENCES
- 1.
- 2.
UFSAR, Chapter 6.
- 3.
- 4.
DPC-NE-3004.
- 5.
10 CFR 50.36, Technical Specification, (c)(2)(ii).
- 6.
WCAP-15049-A, Rev. 1, April 1999.
- 7.
NUREG-1366, February 1990.
- 8. WCAP-18546-P-A, March 2023.
- 9. WCAP-16996-P-A, Rev. 1, November 2016.
ECCS - Operating B 3.5.2 BASES Catawba Units 1 and 2 B 3.5.2-3 Revision No. 5 BACKGROUND (continued)
The high and intermediate head subsystems of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the moderator temperature coefficient is highly negative, such as at the end of each cycle.
During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.
The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a safety injection actuation.
The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet GDC 35 (Ref. 1).
APPLICABLE The LCO helps to ensure that the following acceptance criteria for the SAFETY ANALYSES ECCS, established by 10 CFR 50.46 (Ref. 2), will be met following a small break LOCA and there is a high level of probability that the criteria are met following a large break LOCA:
- a.
Maximum fuel element cladding temperature is d 2200qF;
- b.
Maximum cladding oxidation is d 0.17 times the total cladding thickness before oxidation;
- c.
Maximum hydrogen generation from a zirconium water reaction is d 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; (Ref. 2 and Ref. 8) are met.
ECCS - Operating B 3.5.2 BASES Catawba Units 1 and 2 B 3.5.2-4 Revision No. 5 APPLICABLE SAFETY ANALYSES (continued)
- d.
Core is maintained in a coolable geometry; and
- e.
Adequate long term core cooling capability is maintained.
The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment pressure and temperature limits are met.
Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Refs. 3 and 4). This event has the greatest potential to challenge the limits on runout flow set by the manufacturer of the ECCS pumps. It also sets the maximum response time for their actuation.
Direct flow from the centrifugal charging pumps and SI pumps is credited in a small break LOCA event. The RHR pumps are also credited, for larger small break LOCAs, as the means of supplying suction to these higher head ECCS pumps after the switch to sump recirculation. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The MSLB analysis also credits the SI and centrifugal charging pumps. Although some ECCS flow is necessary to mitigate a SGTR event, a single failure disabling one ECCS train is not the limiting single failure for this transient. The SGTR analysis primary to secondary break flow is increased by the availability of both centrifugal charging and SI trains. Therefore, the SGTR analysis is penalized by assuming both ECCS trains are operable as required by the LCO. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions:
- a.
A large break LOCA event, with loss of offsite power and a single failure disabling one ECCS train; and
- b.
A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.
During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.
The effects on containment mass and energy releases are accounted for in appropriate analyses (Ref. 3). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core being uncovered following a large LOCA.
or without
ECCS - Operating B 3.5.2 BASES Catawba Units 1 and 2 B 3.5.2-11 Revision No. 5 SURVEILLANCE REQUIREMENTS (continued)
Inspections will consist of a visual examination of the exterior surfaces of the strainer assembly for any evidence of debris, structural distress or abnormal corrosion. The intent of this surveillance is to ensure the absence of any condition which could adversely affect strainer functionality. Surveillance performance does not require removal of any tophat modules or grating, but the strainer exteriors shall be visually inspected. This surveillance is not a commitment to inspect 100% of the surface area of all tophats, but a sufficiently detailed inspection of exterior strainer surfaces is required to establish a high confidence that no adverse conditions are present. The scope of inspection necessary to provide high confidence includes 100% of the strainer areas that can be accessed and inspected using normal means and tools (i.e., flashlight, extendable mirror, hand held digital camera) without disassembly, and that difficult to access areas will be inspected to the extent possible using these same means.
Any damage detected in the strainer assembly inspection will result in an expansion of the scope of the inspection to include other areas of potential damage. Inspection scope should be expanded, as needed, for degradation of strainer components identified during this inspection that were not considered readily accessible during the inspectors initial evaluation.
REFERENCES
- 1.
10 CFR 50, Appendix A, GDC 35.
- 2.
- 3.
UFSAR, Section 6.2.1.
- 4.
UFSAR, Chapter 15.
- 5.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 6.
NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.
- 7.
IE Information Notice No. 87-01.
- 8. WCAP-18546-P-A, March 2023.
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WCAP-16996-P-A, Rev. 1, November 2016. (Unit 1)
WCAP-12945-P-A, March 1998. (Unit 2)
Containment Spray System B 3.6.6 BASES Catawba Units 1 and 2 B 3.6.6-3 Revision No. 8 APPLICABLE The limiting DBAs considered relative to containment OPERABILITY SAFETY ANALYSES are the loss of coolant accident (LOCA) and the steam line break (SLB).
The DBA LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients.
No two DBAs are assumed to occur simultaneously or consecutively.
The postulated DBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of the Containment Spray System, the RHR System, and the ARS being rendered inoperable (Ref. 2).
The DBA analyses show that the maximum peak containment pressure results from the LOCA analysis, and is calculated to be less than the containment design pressure. The maximum peak containment atmosphere temperature results from the SLB analysis and was calculated to be within the containment environmental qualification temperature during the DBA SLB. The basis of the containment environmental qualification temperature is to ensure the OPERABILITY of safety related equipment inside containment (Ref. 3).
The Containment Spray System actuation modeled in the containment analysis is based on the time associated with reaching the RWST low level setpoint prior to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time is composed of operator action delay and system startup time.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 4).
Inadvertent actuation is precluded by a design feature consisting of an additional set of containment pressure sensors which prevents operation when the containment pressure is below the containment pressure control system permissive.
The Containment Spray System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 5).
the approved method
Containment Spray System B 3.6.6 BASES Catawba Units 1 and 2 B 3.6.6-8 Revision No. 8 SURVEILLANCE REQUIREMENTS (continued)
Accumulated gas should be eliminated or brought within the acceptance criteria limits.
Containment Spray System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.
REFERENCES
- 1.
10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 2.
UFSAR, Section 6.2.
- 3.
- 4.
- 5.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 6.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
WCAP-12945-P-A, March 1998. (Unit 2)
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U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 3 ATTACHMENT 3 PLANT OPERATING PARAMETERS COMPARED TO TECHNICAL SPECIFICATION LIMITS 2 PAGES PLUS THE COVER
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 3 Plant Operating Parameters Compared to Technical Specification Limits Precedent submittals listed in Section 4.2 of Enclosure 1 addressing the adoption of the FSLOCA EM were reviewed to examine requests for additional information (RAIs) that were issued by the NRC staff. The information presented below focuses on an RAI related to how FSLOCA plant input parameters correspond to Technical Specification (TS) Limiting Condition for Operation (LCO) values. Specifically, it was requested that a comparison be provided for the plant operating parameters used in the FSLOCA analysis to the TS limits for the same parameters, where applicable.
Limitation and Condition 11 of the FSLOCA Evaluation Methodology states, 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.
Duke Energy is not proposing any changes to the CNS or MNS TS LCOs as part of this license amendment request. Table 1 below provides a comparison of the plant operating parameters used in the FSLOCA analysis from Tables 1 and 2 of Enclosures 2 (proprietary) and 4 (non-proprietary), versus the corresponding current TS LCO limits.
Table 1: Plant Operating Parameters Compared to the Technical Specification Limits for Catawba Unit 1 and McGuire Units 1 & 2 Parameter As-Analyzed Value or Range for AXIOM clad fuel TS Limit LCO Number Core power 3700 MWt + 0.3%
calorimetric uncertainty, 3711.1 MWt uncertainty adjusted 3469 MWt Rated Thermal Power CNS/MNS TS 1.1 Heat Flux Hot Channel Factor FQ(Z) 2.7 (maximum value, with burndown effects included)
Included in the Core Operating Limits Report (COLR)
CNS/MNS TS 3.2.1 Nuclear Enthalpy Rise Hot Channel Factor FH 1.72 (maximum value, with burndown effects included)
Included in the COLR CNS/MNS TS 3.2.2 Axial flux difference band at Full Power
-22.58% / +14.58%
Included in the COLR CNS/MNS TS 3.2.3 Low pressurizer pressure reactor trip setpoint 1800 psig 1938 psig (CNS) 1935 psig (MNS)
CNS/MNS TS 3.3.1 Low pressurizer pressure safety injection actuation setpoint 1700 psig 1839 psig (CNS) 1835 psig (MNS)
CNS/MNS TS 3.3.2
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 3 Minimum RCS Flow Rate 379,464 gpm 384,000 gpm (CNS -
Unit 1) 388,000 gpm (MNS)
CNS/MNS TS 3.4.1 Accumulator water volume 6790 VACC 7422 gal; 907.7 VACC 992.2 ft3 (CNS) 7630 VACC 8079 gal, 1020 VACC 1080 ft3; (MNS) 6870 VACC 7342 gal, 918.4 VACC 981.5 ft3 CNS/MNS TS 3.5.1 Accumulator gas cover pressure 555 PACC 669 psig (Region I),
555 PACC 708 psig (Region II);
585 PACC 678 psig (CNS) 585 PACC 639 psig (MNS)
CNS/MNS TS 3.5.1 Minimum accumulator water boron concentration 1950 ppm Included in the COLR CNS/MNS TS 3.5.1 Safety Injection (SI) water temperature 60°F SI Temp 110°F; 70°F Temp 100°F RWST CNS/MNS TS 3.5.4 Main Steam Safety Valves (MSSV) opening setpoint for second stage 1190 psig + 14.7 =
1204.7 psia.
1.03*1204.7 = 1240.8 psia.
Lift Setting (psig + 3%)
CNS MNS 1190 1190 CNS/MNS TS 3.7.1 Minimum initial containment pressure
-0.3 psig (14.4 psia)
-0.1 to +0.3 psig (CNS)
-0.3 to +0.3 psig (MNS)
CNS/MNS TS 3.6.4 Maximum initial upper containment temperature 105 °F 75 to 100 °F CNS/MNS TS 3.6.5 Maximum initial lower containment temperature 125 °F 100 to 120 °F CNS/MNS TS 3.6.5 Minimum air return fan (deck fan) delay time 8 minutes 8 to 10 minutes CNS/MNS TS 3.6.11 Maximum initial ice bed temperature 30 °F 27 °F CNS/MNS TS 3.6.12 and TS 3.6.13
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 4 ATTACHMENT 4 COMPARISON OF CQD AND FSLOCA INPUT PARAMETERS 3 PAGES PLUS THE COVER
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 4 Comparison of CQD and FSLOCA Input Parameters Precedent submittals listed in Section 4.2 of Enclosure 1 addressing the adoption of the FSLOCA EM were reviewed to examine requests for additional information (RAIs) that were issued by the NRC staff. The information presented below focuses on an RAI related to how FSLOCA plant input parameters may differ from previous inputs used in existing Large Break LOCA analyses. Specifically, it was requested that a comparison be provided for the input parameters used in the FSLOCA analysis and the UFSAR parameter values, and provide justification for the differences.
The current Code Qualification Document (CQD) Best Estimate Large Break LOCA analyses were implemented at CNS/MNS in 2000, where one composite bounding analysis was applied to all four CNS/MNS units. The FSLOCA Evaluation Methodology (EM) is a new best-estimate method that incorporates new conservatisms requiring a host of new inputs. 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 methodology. Plant operating ranges considered in the CQD Large Break LOCA analysis are described in CNS UFSAR Table 15-71 and MNS UFSAR Table 15-55. Initial conditions used for the CQD minimum post-LOCA containment pressure are described in CNS UFSAR Table 6-66 and MNS UFSAR Tables 6-64 and 6-65. FSLOCA inputs, including those that were changed relative to the CQD Large Break LOCA analysis, are discussed below and shown in Table 2.
Core thermal power was increased in the FSLOCA analyses to bound an Extended Power Uprate (EPU) to 3700 MWt for MNS Units 1 & 2, and CNS U1. The 0.3% core thermal power uncertainty was included such that the power used in the analysis was 3711.1 MWt. Core peaking factors were adjusted to provide margin to core designs for AXIOM clad fuel.
Accumulator parameters were selected to bound an FSLOCA composite model approach for MNS Units 1 & 2 and CNS U1, as described in Section 3 of Enclosures 2 (proprietary) and 4 (non-proprietary). This is the same approach that was used for accumulator parameters in the CQD analysis.
FSLOCA RCS flow rate is based on MNS Unit 1 Thermal Design flow of 379,464 gpm, which is based on use of MNS Unit 1 vessel parameters used in the composite model approach for MNS Units 1 & 2, and CNS Unit 1. The CQD Large Break analysis was originally performed using a total RCS flow rate of 390,000 gpm. A subsequent CQD assessment was performed for a reduction in total RCS flow to 379,464 gpm Thermal Design Flow with a 0°F impact to the CQD analysis. This assessment was described in a letter to the NRC dated April 9, 2003 (ADAMS Accession No. ML031060339) via annual 10 CFR 50.46 reporting.
Minimum injected ECCS flows assumed in CQD LBLOCA analyses are described in CNS UFSAR Table 15-65 and MNS UFSAR Table 15-45. In 2006, the Small Break LOCA was re-analyzed to support decreased intermediate head safety injection (IHSI) and high head safety injection (HHSI) pump flows at all MNS/CNS units. The reduced flows were used in the Small Break LOCA analyses to account for the TS surveillance limits of +/- 2% on the frequency of the emergency diesel generators (EDGs), and to obtain additional test acceptance criteria margin for the IHSI and HHSI pumps. The CQD Large Break LOCA analyses were not re-analyzed as sufficient low head safety injection (LHSI) flow margin exists in the current analyses to offset the reduced flow due to the frequency reduction of the EDGs. This assessment was described in a letter to the NRC dated May 22, 2007 (ADAMS Accession No. ML071500297) via 30-Day 10 CFR 50.46 reporting.
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 4 Since CNS and MNS have implemented the Water Management Initiative, the transition to cold leg recirculation is delayed since the containment spray pumps do not auto-start in the injection phase. Containment spray pumps are manually started after the RWST reaches the Low Level setpoint. Automatic action of the containment spray system no longer occurs. For the Region II analysis, the analyzed duration is less than or equal to 600 seconds, which is well before the RWST Low Level setpoint would be reached, so containment spray is not actuated for the FSLOCA minimum containment pressure calculation, per Table 2 of Enclosures 2 (proprietary) and 4 (non-proprietary).
Table 2 - Comparison of CQD and FSLOCA Input Parameters Parameter Operating Range, CQD Large Break LOCA; CNS UFSAR Tables 15-71 & 6-66, and MNS UFSAR Tables 6-64, 6-65, & 15-55 Operating Range, FSLOCA for CNS Unit 1 and MNS Units 1 & 2 Reactor Power Core power 100.3% of 3469 MWt, ( 3479.4 MWt)
Core power 100.3% of 3700 MWt ( 3711.1 MWt), bounds EPU conditions Peak Heat Flux Hot Channel peaking factor FQ 2.7 ( 4 ft), 2.5 (> 4 ft);
includes TCD burndown effects FQ 2.70; includes burndown effects Peak Enthalpy Rise Hot Channel peaking factor FH 1.67; includes TCD burndown effects FH 1.72; includes burndown effects Axial power distribution Established per EM Established per EM RCS Tavg 581.1°F Tavg 593.9°F, 587.5° +/- 6.4°F 581.0°F Tavg 594.0°F, 587.5° +/- 6.5°F [Note 1]
Pressurizer pressure 2190 psia PRCS 2310 psia, 2250 psia +/- 60 psi 2200 psia PRCS 2300 psia, 2250 psia +/- 50 psi [Note 1]
RCS Loop flow 97,500 gpm/loop; 390,000 gpm total Subsequently evaluated for reduction to 379,464 gpm total flow.
94,866 gpm/loop for all units; 379,464 gpm total Maximum SG tube plugging level 10% (CNS 2),
5% (MNS and CNS 1) 10% (All Units)
Accumulator fluid temperature 105°F TACC 125°F 105°F TACC 125°F Accumulator pressure 555 psig PACC 708 psig 555 psig PACC 669 psig (Region I) 555 psig PACC 708 psig (Region II);
Accumulator liquid volume 6790 gal VACC 7422 gal (MNS),
7550 gal VACC 8159 gal (CNS) 6790 gal VACC 7422 gal; Minimum accumulator boron concentration 2275 ppm 1950 ppm Minimum Injected ECCS flows CNS UFSAR Table 15-65, MNS UFSAR Table 15-45 Tables 4 and 5 of Enclosures 2 and 4 Safety injection temperature 58°F SI Temp 90°F, covers a RWST temperature range of 70-100°F and component cooling water temperature down to 45°F 60°F SI Temp 110°F;
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 - Attachment 4 Safety injection delay 17 seconds (with OPA) 32 seconds (with LOOP) 17 seconds (with OPA) 32 seconds (with LOOP)
Initial Containment pressure 14.7 psia 14.4 psia Minimum containment outside air / ground temperature N/A
-5°F Maximum Initial Ice Condenser temperature 27°F 30°F Minimum Initial Refueling Water Storage Tank temperature 65°F 58°F used for broken loop spilling flow Maximum containment spray system flow, total 9600 gpm 0 gpm Maximum number of containment spray pumps operating 2
0 Fastest post-LOCA initiation of spray system (assuming off-site power loss at start of LOCA) 25 seconds N/A Fastest post-LOCA initiation of containment air return fans 480 seconds 8 minutes after break Note 1: New value based on updated measurement uncertainty
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 ENCLOSURE 3 WESTINGHOUSE AFFIDAVIT 3 PAGES PLUS THE COVER AXIOM, FULL SPECTRUM, FSLOCA, ZIRLO, and Optimized ZIRLO are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world.
All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
- This record was final approved on 07/23/2025 19:12:06. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-039 Page 1 of 3 Commonwealth of Pennsylvania:
County of Butler:
(1)
I, Jerrod Ewing, Manager, Operating Plants Licensing; Cranberry Township, PA, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).
(2)
I am requesting the proprietary portions of RA-25-0014, Enclosure 2 be withheld from public disclosure under 10 CFR 2.390.
(3)
I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.
(4)
Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.
(ii)
The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.
(iii)
Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
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Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-039 Page 2 of 3 (5)
Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).
(c)
Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
(6)
The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.
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- This record was final approved on 07/23/2025 19:12:06. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-039 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.
Executed on: 7/23/2025 Signed electronically by Jerrod Ewing
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U.S. Nuclear Regulatory Commission Serial: RA-25-0014 ENCLOSURE 4 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO MCGUIRE UNITS 1 AND 2 AND CATAWBA UNIT 1 (NON-PROPRIETARY VERSION)
Note: Proprietary information is identified by bolded brackets and has been removed.
63 PAGES PLUS THE COVER AXIOM, FULL SPECTRUM, FSLOCA, ZIRLO, and Optimized ZIRLO are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world.
All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 2 of 64 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO MCGUIRE UNITS 1 AND 2 AND CATAWBA UNIT 1
1.0 INTRODUCTION
An analysis with the FULL SPECTRUM' loss-of-coolant accident (FSLOCA') evaluation model (EM) has been completed for McGuire Units 1 and 2 and Catawba Unit 1 to support 18-month cycle operation with AXIOM cladding and the PRIME' bottom nozzle. This analysis also supports an extended power uprate (EPU), though the EPU will not be implemented at this time. This license amendment request (LAR) for McGuire Units 1 and 2 and Catawba Unit 1 requests approval to apply the Westinghouse FSLOCA EM.
The FSLOCA EM (Reference 1) was developed to address the full spectrum of loss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM 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 (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as small-break LOCAs (SBLOCAs) and large-break LOCAs (LBLOCAs).
The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as SBLOCAs. Region II includes break sizes that are typically defined as LBLOCAs.
The FSLOCA EM explicitly considers the effects of fuel pellet thermal conductivity degradation (TCD) and other burnup-related effects by calibrating to fuel rod performance data input generated by the PAD5 code (Reference 2), which explicitly models TCD and is benchmarked to high burnup data in Reference 2.
The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet TCD.
The emergency core cooling system (ECCS) acceptance criteria that apply to this analysis are specified for AXIOM cladding in Section 6.2.1.4 of Reference 3. Per Section 6.2.1.4 of Reference 3, the Westinghouse approach used to satisfy the maximum cladding temperature (2,200°F) and maximum hydrogen generation (i.e., core-wide oxidation (CWO)) (1%) acceptance criteria defined in 10 CFR 50.46 (b)(1) and (b)(3)
(Reference 4), respectively, is applicable to AXIOM cladding. However, the maximum local oxidation (MLO) acceptance criterion (17%) defined in 10 CFR 50.46 (b)(2) is replaced with the Nuclear Regulatory Commission (NRC) approved AXIOM cladding performance-based embrittlement acceptance criterion.
The Cathcart-Pawel equivalent cladding reacted (ECR) is confirmed to remain below the ductile-to-brittle transition (DBT) limit for AXIOM cladding described in Section 3.11 of Reference 3 (i.e., minimum ECR margin (MEM) 0%). A high probability statement is developed for the peak cladding temperature (PCT),
MEM, and CWO that is needed to demonstrate compliance with these acceptance criteria via statistical methods.
Section 6.2.1.4 of Reference 3 indicates that the breakaway time is not a plausibly limiting acceptance criterion for the SBLOCA and LBLOCA analyses with AXIOM cladding fuel rods. The coolable geometry acceptance criterion, 10 CFR 50.46 (b)(4), is assured by compliance with the first two acceptance criteria (PCT and local oxidation (MEM)), and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies such that a coolable geometry is maintained. Further, the FSLOCA EM does not address the long-term cooling acceptance criterion defined in 10 CFR 50.46 (b)(5). Per Section 6.2.1.4 of Reference 3, the Westinghouse approach used to satisfy the long-term cooling acceptance criterion remains applicable to AXIOM cladding.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 3 of 64 The FSLOCA EM has been generically approved by the NRC for Westinghouse 3-loop and 4-loop plants with cold leg ECCS injection (Reference 1). Since McGuire Units 1 and 2 and Catawba Unit 1 are Westinghouse designed 4-loop plants with cold leg ECCS injection, the approved method is applicable.
Information required to address Limitations and Conditions 9 and 10 of the NRCs safety evaluation report (SER) for Reference 1 was docketed in Reference 5 in support of application of the FSLOCA EM to Westinghouse 4-loop plants.
This report summarizes the application of the Westinghouse FSLOCA EM to McGuire Units 1 and 2 and Catawba Unit 1. The application of the FSLOCA EM to McGuire Units 1 and 2 and Catawba Unit 1 is consistent with the NRC-approved methodology (Reference 1), as modified for AXIOM cladding in Reference 3, with exceptions identified under Limitation and Condition Number 2 in Section 2.3.
The application of the FSLOCA EM to McGuire Units 1 and 2 and Catawba Unit 1 is consistent with the limitations and conditions as identified in the NRCs SER for Reference 1, and is also applicable for the McGuire Units 1 and 2 and Catawba Unit 1 plant design and operating conditions.
A composite FSLOCA EM analysis was performed to cover McGuire Units 1 and 2 and Catawba Unit 1.
The composite model approach is described in Section 3.0.
Both Duke Energy and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with Westinghouse internal processes for assessing EM changes and errors, are used to identify the need for LOCA analysis impact assessments.
The major plant parameter and analysis assumptions used in the McGuire Units 1 and 2 and Catawba Unit 1 analysis with the FSLOCA EM are provided in Tables 1 through 5.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 4 of 64 2.0 METHOD OF ANALYSIS 2.1 FULL SPECTRUM LOCA Evaluation Model Development In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 (Reference 4 and Reference 6) and Appendix K, ECCS Evaluation Models, to permit the use of a realistic EM to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate thermal-hydraulic models may be used in place of models with Appendix K features. After the rule change, Westinghouse developed and received approval for a best-estimate LBLOCA EM, which is discussed in Reference 7.
The EM is referred to as the Code Qualification Document (CQD), and was developed following Regulatory Guide (RG) 1.157 (Reference 8).
When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles of RG 1.157, while providing a more systematic approach to the development and assessment process of a PWR accident and safety analysis EM. Therefore, the development of the FSLOCA EM followed the Evaluation Model Development and Assessment Process (EMDAP), which is documented in RG 1.203. While RG 1.203 expands upon RG 1.157, there are certain aspects of RG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCA EM.
2.2 WCOBRA/TRAC-TF2 Computer Code The FSLOCA EM (Reference 1) uses the WCOBRA/TRAC-TF2 code to analyze the system thermal-hydraulic response for the full spectrum of break sizes. WCOBRA/TRAC-TF2 was created by combining a 1D module (TRAC-P) with a 3D module (based on Westinghouse modified COBRA-TF).
The 1D and 3D modules include an explicit non-condensable gas transport equation. The use of TRAC-P allows for the extension of a two-fluid, six-equation formulation of the two-phase flow to the 1D loop components. This new code is WCOBRA/TRAC-TF2, where TF2 is an identifier that reflects the use of a three-field (TF) formulation of the 3D module derived by COBRA-TF and a two-fluid (TF) formulation of the 1D module based on TRAC-P.
This best-estimate computer code contains the following features:
1.
Ability to model transient three-dimensional flows in different geometries inside the reactor vessel 2.
Ability to model thermal and mechanical non-equilibrium between phases 3.
Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes 4.
Ability to represent important reactor and plant components such as fuel rods, steam generators (SGs), reactor coolant pumps (RCPs), etc.
A detailed assessment of the computer code WCOBRA/TRAC-TF2 was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena for a LOCA. Modeling of a LOCA introduces additional uncertainties which are identified and quantified in the plant-specific analysis. The reactor vessel and loop noding scheme used in the FSLOCA EM is consistent with the noding scheme used for the experiment simulations that form the validation basis for the physical models in the code. Such noding choices have been justified by assessing the model against large and full scale experiments.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 5 of 64 2.3 Compliance with FSLOCA EM Limitations and Conditions The NRCs SER for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. A summary of each limitation and condition and how it was met is provided below.
Limitation and Condition Number 1 Summary The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.
Compliance 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 Section 6.0 and is not being used to demonstrate compliance with 10 CFR 50.46 (b)(5).
Limitation and Condition Number 2 Summary The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.
Compliance 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 pursuant to 10 CFR 50.46 in References 10 through 16 and the changes for applications of AXIOM cladding, as described in Reference 3.
The analysis was performed with a code version which incorporated the changes and error corrections described in References 10 through 16, except for the error in the steam/fission gas specific heat calculation described in Reference 16. This error was found to have a negligible impact on analysis results with the FSLOCA EM, leading to an estimated PCT impact of 0°F, as described in Reference 16.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 6 of 64 Limitation and Condition Number 3 Summary For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled).
This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.
Compliance 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.
Limitation and Condition Number 4 Summary The decay heat uncertainty multiplier will be [
]a,c The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.
Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was [
]a,c 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.
Limitation and Condition Number 5 Summary The maximum assembly and rod length-average burnup is limited to [
]a,c respectively.
Compliance The maximum analyzed assembly and rod length-average burnup were less than or equal to [
]a,c respectively, for McGuire Units 1 and 2 and Catawba Unit 1.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 7 of 64 Limitation and Condition Number 6 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PAD5 code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.
Compliance 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, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 2.
Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter should be [
]a,c Compliance Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was [
]a,c for the McGuire Units 1 and 2 and Catawba Unit 1 Region I analysis.
Limitation and Condition Number 8 Summary The [
]a,c Compliance Consistent with the NRC-approved methodology, the [
]a,c for the McGuire Units 1 and 2 and Catawba Unit 1 Region I analysis.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 8 of 64 Limitation and Condition Number 9 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the [
]a,c for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.
Compliance 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.
Limitation and Condition Number 10 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to:
- 1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and 2) ensure that the [
]a,c must cover the equivalent 2 to 4-inch break range using RCS-volume scaling relative to the demonstration plant.
This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.
Additionally, the minimum sampled break area for the analysis of Region II should be 1 ft2.
Compliance 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.
The minimum sampled break area for the McGuire Units 1 and 2 and Catawba Unit 1 Region II analysis is 1 ft2.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 9 of 64 Limitation and Condition Number 11 Summary There are various aspects of this Limitation and Condition, which are summarized below:
- 1. The [
]a,c the Region I and Region II analysis seeds, and the analysis inputs will be declared and documented prior to performing the Region I and Region II uncertainty analyses. The [
]a,c and the Region I and Region II analysis seeds will not be changed throughout the remainder of the analysis once they have been declared and documented.
- 2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT, MEM, and CWO which caused the input changes will be provided.
These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.
- 3. Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.
Compliance This Limitation and Condition was met for the McGuire Units 1 and 2 and Catawba Unit 1 analysis as follows:
- 1. The [
]a,c 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 [
]a,c 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.
Limitation and Condition Number 12 Summary The plant-specific dynamic pressure loss from the SG secondary side to the main steam safety valves (MSSVs) must be adequately accounted for in analysis with the FSLOCA EM.
Compliance A bounding plant-specific dynamic pressure loss from the 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, [
]a,c 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.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 10 of 64 Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM: 1) the [
]a,c and 2) the
[
]a,c Compliance The [
]a,c in the analysis for McGuire Units 1 and 2 and Catawba Unit 1. The [
]a,c in the analysis.
Limitation and Condition Number 14 Summary For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation acceptance criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.
Compliance 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 ECR is confirmed to remain below the DBT limit for AXIOM cladding described in Section 3.11 of Reference 3. Limitation and Condition Number 14 is therefore not applicable to the McGuire Units 1 and 2 and Catawba Unit 1 FSLOCA EM analysis.
Limitation and Condition Number 15 Summary The Region II analysis will be executed twice; once assuming offsite power available (OPA) and once assuming loss-of-offsite power (LOOP). The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.
The [
]a,c Compliance 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 [
]a,c
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 11 of 64 3.0 COMPOSITE MODEL APPROACH The current licensing basis LBLOCA analysis for McGuire Units 1 and 2 and Catawba Units 1 and 2 was performed with the CQD EM. The composite model approach was described in References 18 and 19, and the composite analysis was approved by the NRC in References 20 and 21. The same approach was used in the FSLOCA EM analysis for McGuire Units 1 and 2 and Catawba Unit 1, except for the changes noted below.
The composite FSLOCA EM analysis does not cover Catawba Unit 2. A separate FSLOCA EM analysis will be performed for Catawba Unit 2 due to the different SG design and the importance of the SG design to the progression of Region I transients (see Section 2.3.2.7 of Reference 1).
As described on pages 1 and 2 of the Attachment to Reference 18, two vessel models were built to capture the differences in the upper internals among the units. Other minor differences were bounded in the two vessel models. Sensitivity studies were performed to identify the limiting vessel model. Page 50 of the presentation in Enclosure 2 of Reference 19 indicates that the McGuire Unit 1 vessel model was determined to be limiting, so this model was used in the CQD LBLOCA analysis. It was confirmed that the McGuire Unit 1 vessel model is also appropriate for the Region I and Region II analyses with the FSLOCA EM, so this vessel model was also used in these analyses.
As described on page 2 of the Attachment to Reference 18, sensitivity studies were performed to identify the direction of conservatism for the accumulator line resistance and accumulator water volume. Pages 32 and 41 of the presentation in Enclosure 2 of Reference 19 indicate that minimum line resistance and minimum water volume were determined to be limiting. Therefore, the minimum line resistance (based on McGuire Unit 2) and the minimum water volume range (based on McGuire Units 1 and 2) were used in the CQD LBLOCA analysis. The wider accumulator pressure range for Catawba Units 1 and 2, which encompasses the range for McGuire Units 1 and 2, was also used.
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 same accumulator modeling approach used in the CQD LBLOCA analysis was also used in the Region II analysis with the FSLOCA EM (minimum line resistance, minimum water volume range, wider pressure range).
Figure 31.3-4 of Reference 1 shows a correlation between accumulator pressure and PCT for the Region I demonstration analysis with the FSLOCA EM. Lower accumulator pressure delays the accumulator injection which delays the termination of the boil-off heatup for SBLOCA transients, leading to a higher PCT. The wider accumulator pressure range (555 - 708 psig) was used in the CQD LBLOCA analysis and was also used in the Region II analysis with the FSLOCA EM. The narrower accumulator pressure range for McGuire Units 1 and 2 (555 - 669 psig) results in lower sampled pressures and thus higher PCTs for Region I, so the narrower accumulator pressure range was used in the Region I analysis with the FSLOCA EM for McGuire Units 1 and 2 and Catawba Unit 1. The 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).
Reference 22 describes the ongoing process that has been followed to assure that the composite CQD analysis for McGuire Units 1 and 2 and Catawba Units 1 and 2 has remained representative and bounding for these units. The same process will be followed to assure that the composite FSLOCA EM analysis for McGuire Units 1 and 2 and Catawba Unit 1 remains representative and bounding for these units.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 12 of 64 4.0 REGION I ANALYSIS 4.1 Description of Representative Transient The small-break LOCA transient can be divided into time periods in which specific phenomena are occurring, as discussed below.
Blowdown The rapid depressurization of the RCS coincides with subcooled liquid flow through the break. Following the reactor trip on the low pressurizer pressure setpoint, the pressurizer drains, and safety injection (SI) is initiated on the low pressurizer pressure SI setpoint. After reaching this setpoint and applying the SI delays, high head safety injection (HHSI) flow begins. Phase separation begins in the upper head and upper plenum near the end of this period until the entire RCS eventually reaches saturation, ending the rapid depressurization slightly above the SG secondary side pressure near the modeled MSSV setpoint.
Natural Circulation This quasi-equilibrium phase persists while the RCS pressure remains slightly above the secondary side pressure. The system drains from the top down, and while significant mass is continually lost through the break, the vapor generated in the core is trapped in the upper regions of the primary side (reactor vessel, hot legs, and steam generators) by the liquid remaining in the crossover leg loop seals. Throughout this period, the core remains covered by a two-phase mixture and the fuel cladding temperatures remain at the saturation temperature level.
Loop Seal Clearance As the system drains, the liquid levels in the downhill side of the pump suction (crossover leg) become depressed all the way to the bottom elevations of the piping, allowing the steam trapped during the natural circulation phase to vent to the break (i.e., a process called loop seal clearance). The break flow and the flow through the RCS loops with a cleared loop seal become primarily vapor. Relief of a static head imbalance allows for a quick but temporary recovery of liquid levels in the inner portion of the reactor vessel.
Boil-Off With a vapor vent path established after the loop seal clearance, the RCS depressurizes at a rate controlled by the critical flow, which continues to be a primarily high-quality mixture of water and steam. The RCS pressure remains high enough such that SI flow cannot make up for the primary system fluid inventory lost through the break, leading to core uncovery and a fuel rod cladding temperature heatup.
Core Recovery The RCS pressure continues to decrease, and once it reaches that of the accumulator gas pressure, the introduction of additional ECCS water from the accumulators replenishes the reactor vessel inventory and recovers the core mixture level. The transient is considered over as the break flow is compensated by the SI flow.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 13 of 64 4.2 Analysis Results The McGuire Units 1 and 2 and Catawba Unit 1 Region I analysis was performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. [
]a,c The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Item 5.0a in Table 1. Control rod drop is modeled for breaks less than 1 square foot assuming a 2.0-second signal delay time and a 3.3-second rod drop time. RCP trip is modeled coincident with reactor trip on the low pressurizer pressure setpoint for LOOP transients. When the low pressurizer pressure SI setpoint is reached, there is a delay to account for emergency diesel generator start-up, filling headers, etc., after which SI is initiated into the RCS.
The results of the McGuire Units 1 and 2 and Catawba Unit 1 Region I uncertainty analysis are summarized in Table 6. The sampled decay heat uncertainty multipliers for the Region I analysis cases are provided in Table 16.
Table 7 contains a sequence of events for the transient that produced the Region I analysis PCT result.
Figures 1 through 13 illustrate the calculated key transient response parameters for this transient. Table 8 contains a sequence of events for the transient that produced the Region I analysis MEM result. The CWO result in Table 6 is 0.00%. This result applies to all runs in the Region I analysis, so the sequence of events table is not provided for a specific run with this CWO result.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 14 of 64 5.0 REGION II ANALYSIS 5.1 Description of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occurring.
A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below.
Blowdown - Critical Heat Flux (CHF) Phase In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), and the cladding rapidly heats up and the reactor is shut down due to the core voiding.
The regions of the RCS with the highest initial temperatures (upper core, upper plenum, and hot legs) begin to flash during this period. This phase is terminated when the water in the lower plenum and downcomer begins to flash. The mixture level swells, and a saturated mixture is pushed into the core by the intact loop RCPs, still rotating in single-phase liquid. As the fluid in the cold leg reaches saturation conditions, the discharge flow rate at the break decreases significantly.
Blowdown - Upward Core Flow Phase Heat transfer is increased as the two-phase mixture is pushed into the core. The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.
Blowdown - Downward Core Flow Phase The break flow begins to dominate and pulls flow down through the core as the RCP head degrades due to increased voiding, while liquid and entrained liquid flows also provide core cooling. Heat transfer in this period may be enhanced by liquid flow from the upper head. Once the system has depressurized to less than the accumulator cover pressure, the accumulators begin to inject cold water into the cold legs. During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last period of the blowdown phase.
During this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.
Refill Phase The core continues to heat up as the lower plenum refills with ECCS water. This phase is characterized by a rapid increase in fuel cladding temperature at all elevations due to the lack of liquid and steam flow in the core region. The water completely refills the lower plenum and the refill phase ends. As ECCS water enters the core, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 15 of 64 Reflood Phase During the early reflood phase, the accumulators begin to empty and nitrogen is discharged into the RCS.
The nitrogen surge forces water into the core, which is then evaporated, causing system re-pressurization and a temporary reduction of pumped ECCS flow; this re-pressurization is illustrated by the increase in RCS pressure. During this time, core cooling may increase due to vapor generation and liquid entrainment, but conversely the early reflood pressure spike results in loss of mass out through the broken cold leg.
The pumped ECCS water aids in the filling of the downcomer throughout the reflood period. As the quench front progresses further into the core, the PCT elevation moves increasingly higher in the fuel assembly.
As the transient progresses, continued injection of pumped ECCS water refloods the core, effectively removes the reactor vessel metal mass stored energy and core decay heat, and leads to an increase in the reactor vessel fluid mass. Eventually the core inventory increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.
5.2 Analysis Results The McGuire Units 1 and 2 and Catawba Unit 1 Region II analysis was performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analysis was performed assuming both OPA and LOOP, and the results of both of the OPA and LOOP analyses are compared to the applicable ECCS acceptance criteria. The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Item 5.0a in Table 1. The results of the McGuire Units 1 and 2 and Catawba Unit 1 Region II OPA and LOOP uncertainty analyses are summarized in Table 6. The sampled decay heat uncertainty multipliers for the Region II analysis cases are provided in Table 16.
Table 9 identifies the break size for the cases that produced the analysis results. Tables 10 through 15 contain a sequence of events for these transients. Figures 14 through 27 illustrate the key response parameters for the transient that produced the analysis PCT result for the uncertainty analysis with OPA.
The containment pressure is calculated using the LOTIC2 code (References 23 and 24) for ice condenser containments. The containment model input is summarized in Tables 2 and 3. The conservatively low containment pressure response used for the Region II analysis is compared to the calculated containment backpressure in Figure 21, consistent with the methodology in Reference 1.
Figures 28 and 29 show the PCT versus effective break area multiplier for the Region II OPA and LOOP uncertainty analyses, respectively. These figures reflect the combined effect of the break size and break flow model uncertainties. Figures 30 and 31 show the transient ECR versus PCT for the Region II OPA and LOOP uncertainty analyses, respectively. Figures 32 and 33 show the CWO versus PCT for the Region II OPA and LOOP uncertainty analyses, respectively. Strong trends of increasing ECR and CWO with increasing PCT occur due to the temperature dependence of the oxidation kinetics.
The uncertainty analysis methodology used in the FSLOCA EM is described in Section 30 of Reference 1.
A Monte Carlo sampling of all uncertainty contributors leads to the generation of a sample of simulated results from which upper tolerance limits are derived for the analysis figures of merit (PCT, MEM, CWO).
[
]a,c
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 16 of 64 6.0 COMPLIANCE WITH APPLICABLE ECCS ACCEPTANCE CRITERIA It must be demonstrated that there is a high level of probability that the following acceptance criteria in 10 CFR 50.46, as modified for AXIOM cladding in Section 6.2.1.4 of Reference 3, are met:
(b)(1) The FSLOCA EM analysis PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCTs in Table 6 for McGuire Units 1 and 2 and Catawba Unit 1 are less than 2,200°F, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., Peak Cladding Temperature does not exceed 2,200°F, is demonstrated.
(b)(2) The maximum MLO acceptance criterion (17%) defined in 10 CFR 50.46 (b)(2) is replaced with the NRC-approved AXIOM cladding performance-based embrittlement acceptance criterion.
The FSLOCA EM analysis MEM corresponds to a bounding estimate of the 95th percentile MEM at the 95-percent confidence level. Since the resulting Cathcart-Pawel ECRs for McGuire Units 1 and 2 and Catawba Unit 1 remain below the DBT limit for AXIOM cladding described in Section 3.11 of Reference 3 (i.e., MEM 0%) per Table 6, the analysis confirms that the AXIOM cladding performance-based embrittlement acceptance criterion is demonstrated.
(b)(3) The FSLOCA EM analysis CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWOs in Table 6 for McGuire Units 1 and 2 and Catawba Unit 1 are less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., Core-Wide Oxidation does not exceed 1 percent, is demonstrated.
(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains in a coolable geometry.
This acceptance criterion is met by demonstrating compliance with the first two acceptance criteria (PCT and local oxidation (MEM)), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. The PCT and local oxidation (MEM) acceptance criteria have been met for McGuire Units 1 and 2 and Catawba Unit 1 per Table 6.
It is discussed in Section 32.1 of the NRC-approved FSLOCA EM (Reference 1) that 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 FSLOCA 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. As described in Section 4.2.1.3.2 of the McGuire Nuclear Station Updated Final Safety Analysis Report (UFSAR) and Section 4.2.4.5 of the Catawba Nuclear Station UFSAR, 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, inboard grid deformation due to combined LOCA and seismic loads is not calculated to occur for McGuire Units 1 and 2 and Catawba Unit 1.
(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS.
Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM (Reference 1).
Based on the above discussion, it is concluded that McGuire Units 1 and 2 and Catawba Unit 1 comply with the acceptance criteria in 10 CFR 50.46, as modified for AXIOM cladding in Section 6.2.1.4 of Reference 3.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 17 of 64 7.0 ANALYSIS APPLICABILITY The FSLOCA EM analysis for McGuire Units 1 and 2 and Catawba Unit 1 supports 18-month cycle operation with AXIOM cladding, the PRIME bottom nozzle, and an EPU. This analysis does not support the current cladding material (Optimized ZIRLO' cladding).
During fuel transitions, co-resident fuel effects typically require evaluation. For example, differences in fuel assembly hydraulic resistances can lead to a LOCA penalty. In the case of transitioning to AXIOM cladding and the PRIME bottom nozzle from legacy Westinghouse fuel products, there are no (or negligible) co-resident fuel effects.
The small differences in the bottom nozzle metal mass and pressure loss coefficient have a negligible effect on LOCA analyses. Furthermore, the bottom nozzle loss coefficient is not explicitly modeled in the WCT-TF2 model. The pressure loss information that is used for the WCT-TF2 steady-state calibration, as prescribed by the NRC-approved FSLOCA methodology (Section 26.4 of Reference 1), captures the change in the bottom nozzle loss coefficient. Because the bottom nozzle differences are only considered in the WCT-TF2 steady-state calibration and not in the WCT-TF2 model itself, the PRIME bottom nozzle is considered to be hydraulically similar to the prior bottom nozzle with respect to mixed core effects within the established Westinghouse mixed core evaluation technique. The cladding material change also does not introduce any hydraulic differences. Therefore, a transition core evaluation for the LOCA analysis is not needed to address a fuel assembly hydraulic resistance mismatch.
The change from Optimized ZIRLO cladding to AXIOM cladding leads to differences in the fuel pellet temperatures and rod internal pressures, so separate LOCA analyses are required to cover both cladding materials. During the two transition cycles for each unit, the new fuel product (AXIOM cladding) will be covered by the FSLOCA EM analysis, while the existing fuel product (Optimized ZIRLO cladding) will continue to be covered by the current licensing basis LOCA analyses with the NOTRUMP and CQD EMs.
For the transition cycles, the portions of the Technical Specifications (TS) that apply to the current licensing basis LOCA analyses will be maintained, while the TS changes to support the new fuel/methodologies will also be incorporated. Following discharge of all fuel with Optimized ZIRLO cladding, the FSLOCA EM analysis will cover all fuel in the core, and the portions of the TS that apply to the current licensing basis LOCA analyses will no longer be applicable.
The FSLOCA EM analysis supports the uprated power level (3700 MWt + 0.3% uncertainty per Item 1.0a in Table 1), while the current licensing basis LOCA analyses support the current power level (maximum power, including uncertainty = ~3479 MWt). The current power level cannot be exceeded during the fuel product transition. The uprated power level cannot be used until all fuel with Optimized ZIRLO cladding has been discharged and the FSLOCA EM analysis covers all fuel in the core.
It is noted that maintaining the current licensing basis for the outgoing fuel is consistent with the licensing precedent for vendor-to-vendor fuel transitions. For example, Reference 25 documents the license amendment for the transition to Westinghouse fuel for Sequoyah. Westinghouse performed licensing basis analyses (e.g., PAD5, FSLOCA EM) that were consistent with current standards of the NRC, and the Framatome licensing basis analyses were maintained for their fuel product. The outgoing fuel in the core was covered under the existing licensing basis analyses, which was deemed to be acceptable by the NRC, and only the Westinghouse fuel was explicitly covered under the new analyses.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 18 of 64
8.0 REFERENCES
- 1. Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), WCAP-16996-P-A, Revision 1, November 2016.
- 2. Westinghouse Performance Analysis and Design Model (PAD5), WCAP-17642-P-A, Revision 1, November 2017.
- 3. Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, WCAP-18546-P-A, Revision 0, March 2023.
- 4. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974.
- 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.
- 6. Emergency Core Cooling Systems: Revisions to Acceptance Criteria, Federal Register, V53, N180, pp. 35996-36005, September 1988.
- 7. Code Qualification Document for Best Estimate LOCA Analysis, WCAP-12945-P-A Volume 1, Revision 2 and Volumes 2-5, Revision 1, March 1998.
- 8. Best Estimate Calculations of Emergency Core Cooling System Performance, Regulatory Guide 1.157, USNRC, May 1989.
- 9. Transient and Accident Analysis Methods, Regulatory Guide 1.203, USNRC, December 2005.
- 10. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, LTR-NRC-18-30, Revision 0, July 2018.
- 11. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, LTR-NRC-19-6, Revision 0, February 2019.
- 12. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2020, LTR-NRC-21-5, Revision 0, March 2021.
- 13. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2021, LTR-NRC-22-8, Revision 0, February 2022.
- 14. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2022, LTR-NRC-23-5, Revision 0, March 2023.
- 15. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2023, LTR-NRC-24-6, Revision 0, March 2024.
- 16. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2024, LTR-NRC-25-11, Revision 0, March 2025.
- 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 and 137-139 (Proprietary/Non-Proprietary), Project 700, TAC No. ME5244, LTR-NRC-14-4, Revision 0, January 2014.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 19 of 64
- 18. Duke Energy Corporation, Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, McGuire Nuclear Station, Units 1 and 2, Docket Number 50-369 and 50-370, Implementation of Best-Estimate Large Break LOCA Methodology, April 2000. (ADAMS Accession No. ML003704107)
- 19. MEETING
SUMMARY
- MEETING OF JUNE 12, 2000, REGARDING WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY, USNRC, June 2000. (ADAMS Accession No. ML003728321)
- 20. MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MA8696 AND MA8697), USNRC, September 2000. (ADAMS Accession No. ML003753895)
- 21. CATAWBA NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MA8719 AND MA8720), USNRC, October 2000. (ADAMS Accession No. ML003756631)
- 22. Duke Energy Corporation, McGuire Nuclear Station, Units 1 and 2, Docket Numbers 50-369 and 50-370, Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414, License Amendment Request, Implementation of Best-Estimate Large Break Loss of Coolant Accident (BELBLOCA) Analysis Methodology, August 2000. (ADAMS Accession No. ML003741728)
- 23. Westinghouse Emergency Core Cooling System Evaluation Model - Summary, WCAP-8339, Revision 0, June 1974.
- 24. Long Term Ice Condenser Containment Code - LOTIC Code, WCAP-8354-P-A, Supplement 1, Revision 0, April 1976.
- 25. SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 356 AND 349 REGARDING THE TRANSITION TO WESTINGHOUSE ROBUST FUEL ASSEMBLY-2 (RFA-2) FUEL (EPID L-2020-LLA-0216), USNRC, October 2021. (ADAMS Accession No. ML21245A267)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 20 of 64 Table 1. Plant Operating Range Analyzed and Key Parameters for McGuire Units 1 and 2 and Catawba Unit 1 Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power 3700 MWt + 0.3% uncertainty b) Fuel type 17x17 RFA-2 fuel with AXIOM cladding, PRIME bottom nozzle, and integral fuel burnable absorber (IFBA). Off-spec enriched uranium product (OSEUP) may also be used.
c) Maximum total core peaking factor (FQ),
including uncertainties 2.70 d) Maximum hot channel enthalpy rise peaking factor (FH), including uncertainties 1.72 e) Axial flux difference (AFD) band at 100% power
-22.58% to +14.58%
2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 94,866 gpm/loop b) Vessel average temperature (TAVG) 581.0°F TAVG 594.0°F Coastdown to 572.5°F c) Pressurizer pressure (PRCS) 2200 psia PRCS 2300 psia d) Reactor coolant pump (RCP) model and power Model 93A, 7000 hp 3.0 Containment Parameters a) Containment modeling Region I: Constant pressure equal to initial containment pressure Region II: Conservatively low containment pressure calculated using the information in Tables 2 and 3 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level 10%
b) Main steam safety valve (MSSV) set pressures Pressure at which MSSV begins to open = 1240.8 psia Pressure at which MSSV is fully open = 1343.3 psia These pressures are based on the second stage MSSV, as discussed for Limitation and Condition Number 12 in Section 2.3.
c) Main feedwater temperature 440°F d) Auxiliary feedwater temperature 85°F e) Auxiliary feedwater flow rate 220 gpm (OPA) or 760 gpm (LOOP) (total flow to all SGs)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 21 of 64 Table 1. Plant Operating Range Analyzed and Key Parameters for McGuire Units 1 and 2 and Catawba Unit 1 Parameter As-Analyzed Value or Range 5.0 Safety Injection (SI) Parameters a) Single failure configuration Loss of one train of pumped ECCS b) Safety injection temperature (TSI) 60°F TSI 110°F c) Low pressurizer pressure safety injection safety analysis limit 1700 psig d) Initiation delay time from low pressurizer pressure SI setpoint to full SI flow 17 seconds (OPA) or 32 seconds (LOOP) e) Safety injection flow Minimum flows in Table 4 (Region I) or Table 5 (Region II) 6.0 Accumulator Parameters a) Accumulator temperature (TACC) 105°F TACC 125°F b) Accumulator water volume (VACC) 907.69 ft3 VACC 992.18 ft3 (6790 gal VACC 7422 gal) c) Accumulator pressure (PACC)
Region I Analysis Range: 555 psig PACC 669 psig Region II Analysis Range: 555 psig PACC 708 psig d) Accumulator boron concentration 1950 ppm 7.0 Reactor Protection System Parameters a) Low pressurizer pressure reactor trip signal processing time 2.0 seconds b) Low pressurizer pressure reactor trip setpoint 1800 psig
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 22 of 64 Table 2. Containment Data Used for Region II Calculation of Containment Pressure for McGuire Units 1 and 2 and Catawba Unit 1 Parameter Value Maximum containment upper compartment net free volume 676,255 ft3 Minimum containment lower compartment net free volume 197,800 ft3 Minimum containment dead-ended compartment net free volume 146,052 ft3 Maximum containment upper compartment initial temperature at full power operation 105°F Maximum containment lower compartment initial temperature at full power operation 125°F Maximum containment dead-ended compartment initial temperature at full power operation 125°F Maximum containment ice bed compartment initial temperature at full power operation 30°F Maximum number of containment air return fans (deck fans) in operation during LOCA transient 2 fans Minimum air return fan (deck fan) delay time 8.0 minutes Maximum containment air return fan (deck fan) flow rate per fan 40,000 cfm/fan Maximum number of containment spray pumps in operation during LOCA transient 0
Minimum refueling water storage tank (RWST) temperature for broken loop spilling SI 58°F Active sump maximum volume 76,996 ft3 Containment walls / heat sink properties Table 3 Minimum containment outside air / ground temperature
-5°F Minimum initial containment pressure at normal full power operation 14.4 psia Maximum number of containment venting lines (including purge lines, pressure relief lines or any others) which can be OPEN at onset of transient at full power operation 1 vent line Maximum effective valve diameter of each containment venting line 6.357 inches Maximum delay time between SI signal and start of venting valve closure 28 seconds Maximum venting valve closure time at normal full power operation 5 seconds SI spilling flows 267.1 lbm/s Minimum annulus temperature
-5°F
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 23 of 64 Table 3. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for McGuire Units 1 and 2 and Catawba Unit 1 Wall Area (ft2)
Thickness (ft)
Material 1
21,142 1.34 Concrete 2
5,017 0.0156 1.5 Stainless Steel Concrete 3
24,391 0.058 Carbon Steel 4
31,035 0.0290 Carbon Steel 5
801 0.0625 Stainless Steel 6
57,387 1.97 Concrete 7
9,019 2.04 Concrete 8
3,541 2.5 Concrete 9
2,361 0.0156 1.5 Stainless Steel Concrete 10 768 0.04207 1.5 Carbon Steel Concrete 11 21,278 0.0535 Carbon Steel 12 35,273 0.0535 Carbon Steel 13 14,445 0.0625 Carbon Steel 14 9,040 0.0625 Carbon Steel 15 32,640 0.0026 Stainless Steel 16 51,000 0.00042 Copper 17 9,600 0.00833 Stainless Steel
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 24 of 64 Table 4. Safety Injection Flow Used for Region I Calculation for McGuire Units 1 and 2 and Catawba Unit 1 Pressure (psia)
High Head Safety Injection (HHSI) Flow (gpm/loop)
Intermediate Head Safety Injection (IHSI)
Flow (gpm/loop)
Low Head Safety Injection (LHSI) Flow (gpm/loop) 14.7 89.8 129.6 844.9 34.7 89.0 129.2 801.2 54.7 88.3 128.6 747.3 74.7 87.7 128.1 683.4 94.7 87.1 127.5 609.3 114.7 86.6 126.9 525.0 134.7 86.3 126.2 430.7 154.7 85.3 125.5 326.2 174.7 84.8 124.8 211.5 194.7 84.3 124.0 86.8 194.71 84.3 124.0 0.0 214.7 83.8 123.2 0.0 314.7 81.0 118.6 0.0 414.7 77.9 113.2 0.0 514.7 74.5 106.8 0.0 614.7 70.9 99.6 0.0 714.7 66.9 91.5 0.0 814.7 62.6 82.5 0.0 914.7 58.0 72.6 0.0 1014.7 53.1 61.8 0.0 1114.7 47.9 50.1 0.0 1214.7 42.4 37.6 0.0 1314.7 36.7 24.1 0.0 1414.7 33.7 9.8 0.0 1414.71 33.7 0.0 0.0 1514.7 30.6 0.0 0.0 1614.7 24.1 0.0 0.0 1614.71 0.0 0.0 0.0 2500.0 0.0 0.0 0.0
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 25 of 64 Table 5. Safety Injection Flow Used for Region II Calculation for McGuire Units 1 and 2 and Catawba Unit 1 Pressure (psia)
High Head Safety Injection (HHSI) Flow (gpm/loop)
Intermediate Head Safety Injection (IHSI)
Flow (gpm/loop)
Low Head Safety Injection (LHSI) Flow (gpm/loop) 14.7 89.8 132.0 853.5 34.7 89.0 130.8 700.5 54.7 88.3 129.6 545.8 74.7 87.7 128.5 389.5 94.7 87.1 127.4 231.6 114.7 86.6 126.5 72.1 114.71 86.6 126.5 0.0 134.7 86.3 125.6 0.0 154.7 85.3 121.8 0.0 174.7 84.8 121.0 0.0 194.7 84.3 120.3 0.0 214.7 83.8 119.5 0.0 314.7 81.0 114.7 0.0 414.7 77.9 108.7 0.0 514.7 74.5 101.5 0.0 614.7 70.9 93.1 0.0 714.7 66.9 83.5 0.0 814.7 62.6 72.8 0.0 914.7 58.0 60.8 0.0 1014.7 53.1 47.7 0.0 1114.7 47.9 33.3 0.0 1214.7 42.4 17.8 0.0 1214.71 42.4 0.0 0.0 1314.7 36.7 0.0 0.0 1514.7 30.6 0.0 0.0 1614.7 24.1 0.0 0.0 1614.71 0.0 0.0 0.0 2500.0 0.0 0.0 0.0
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 26 of 64 Table 6. McGuire Units 1 and 2 and Catawba Unit 1 Analysis Results with the FSLOCA EM Outcome Region I Value Region II Value (OPA)
Region II Value (LOOP) 95/95 PCT 1199°F 1662°F 1647°F 95/95 MEM 5.33%
5.07%
4.82%
95/95 CWO 0.00%
0.09%
0.08%
Table 7. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region I Analysis PCT Case Event Time after Break (s)
Start of Transient 0.0 Reactor Trip Signal 16.4 Safety Injection Signal 27.9 Safety Injection Begins 59.9 Loop Seal Clearing Occurs 768 Top of Core Uncovered 1028 Accumulator Injection Begins 1610 PCT Occurs 1611 Top of Core Recovered 1998 Table 8. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region I Analysis MEM Case Event Time after Break (s)
Start of Transient 0.0 Reactor Trip Signal 22.6 Safety Injection Signal 35.5 Safety Injection Begins 67.5 Loop Seal Clearing Occurs 922 Top of Core Uncovered 1136 Accumulator Injection Begins 1436 PCT Occurs 1438 Top of Core Recovered 1602
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 27 of 64 Table 9. Break Sizes for McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis Cases Analysis Case Region II Analysis with OPA Region II Analysis with LOOP Break Type Effective Break Area Multiplier Break Type Effective Break Area Multiplier PCT DEG 2.6613 DEG 2.4245 MEM DEG 2.4586 DEG 2.4586 CWO DEG 2.7980 Split 2.6349 Table 10. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis PCT Case with OPA Event Time after Break (s)
Start of Transient 0.0 Fuel Rod Burst Occurs 2.0 Safety Injection Signal 5.5 Accumulator Injection Begins 12.0 End of Blowdown 22.0 Safety Injection Begins 22.5 Accumulator Empty 55.5 PCT Occurs 112 All Rods Quenched 273
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 28 of 64 Table 11. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis PCT Case with LOOP Event Time after Break (s)
Start of Transient 0.0 Safety Injection Signal 5.4 Fuel Rod Burst Occurs 5.4 PCT Occurs 6.3 Accumulator Injection Begins 11.5 End of Blowdown 15.0 Safety Injection Begins 37.4 Accumulator Empty 53.5 All Rods Quenched 325 Table 12. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis MEM Case with OPA Event Time after Break (s)
Start of Transient 0.0 Fuel Rod Burst Occurs 2.1 Safety Injection Signal 5.4 Accumulator Injection Begins 11.5 End of Blowdown 21.5 Safety Injection Begins 22.4 Accumulator Empty 55.5 PCT Occurs 112 All Rods Quenched 379
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 29 of 64 Table 13. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis MEM Case with LOOP Event Time after Break (s)
Start of Transient 0.0 Fuel Rod Burst Occurs 2.1 Safety Injection Signal 5.4 Accumulator Injection Begins 11.0 End of Blowdown 21.5 Safety Injection Begins 37.4 Accumulator Empty 55.0 PCT Occurs 119 All Rods Quenched 291 Table 14. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis CWO Case with OPA Event Time after Break (s)
Start of Transient 0.0 Fuel Rod Burst Occurs 2.9 Safety Injection Signal 5.1 Accumulator Injection Begins 12.0 End of Blowdown 15.5 Safety Injection Begins 22.1 Accumulator Empty 61.6 PCT Occurs 88.5 All Rods Quenched 358
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 30 of 64 Table 15. Sequence of Events for the McGuire Units 1 and 2 and Catawba Unit 1 Region II Analysis CWO Case with LOOP Event Time after Break (s)
Start of Transient 0.0 Safety Injection Signal 5.3 Accumulator Injection Begins 12.5 End of Blowdown 22.5 Safety Injection Begins 37.3 Fuel Rod Burst Occurs 38.5 Accumulator Empty 54.0 PCT Occurs 87.3 All Rods Quenched 411 Table 16. Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for McGuire Units 1 and 2 and Catawba Unit 1 Region I and Region II Analysis Cases Region Case DECAY_HT (units of )
DECAY_HT (absolute units)*
Region I PCT
+0.3642 2.03%
MEM
+1.0004 5.58%
CWO N/A**
N/A**
Region II (OPA)
+0.1338 0.71%
MEM
+1.2212 6.78%
CWO
+0.6222 3.27%
Region II (LOOP)
+0.5801 3.02%
MEM
+1.2212 6.78%
CWO
+0.4004 2.14%
- Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSI/ANS-5.1-1979 decay heat standard for 235U, 239Pu, and 238U assuming infinite operation.
- No decay heat uncertainty value is provided for the Region I CWO case since the analysis result for all runs is 0.00%.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 31 of 64 Figure 1: McGuire Units 1 and 2 and Catawba Unit 1 Break Flow Void Fraction for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 32 of 64 Figure 2: McGuire Units 1 and 2 and Catawba Unit 1 Total Safety Injection Flow (not including Accumulator Injection Flow) and Total Break Flow for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 33 of 64 Figure 3: McGuire Units 1 and 2 and Catawba Unit 1 RCS Pressure for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 34 of 64 Figure 4: McGuire Units 1 and 2 and Catawba Unit 1 Hot Assembly Two-Phase Mixture Level (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 35 of 64 Figure 5: McGuire Units 1 and 2 and Catawba Unit 1 Peak Cladding Temperature for All Rods for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 36 of 64 Figure 6: McGuire Units 1 and 2 and Catawba Unit 1 Vapor Mass Flow Rate through the Crossover Legs for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 37 of 64 Figure 7: McGuire Units 1 and 2 and Catawba Unit 1 Core Collapsed Liquid Levels (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 38 of 64 Figure 8: McGuire Units 1 and 2 and Catawba Unit 1 Accumulator Injection Flow per Loop for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 39 of 64 Figure 9: McGuire Units 1 and 2 and Catawba Unit 1 Vessel Fluid Mass for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 40 of 64 Figure 10: McGuire Units 1 and 2 and Catawba Unit 1 Steam Generator Secondary Side Pressure for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 41 of 64 Figure 11: McGuire Units 1 and 2 and Catawba Unit 1 Normalized Core Power Shapes for the Region I Analysis PCT Case Note: The localized power decreases occur at grid elevations.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 42 of 64 Figure 12: McGuire Units 1 and 2 and Catawba Unit 1 Relative Core Power for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 43 of 64 Figure 13: McGuire Units 1 and 2 and Catawba Unit 1 Vapor Temperature and Void Fraction at Core Outlet (Hot Assembly Channel) for the Region I Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 44 of 64 Figure 14: McGuire Units 1 and 2 and Catawba Unit 1 Peak Cladding Temperature for All Rods for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 45 of 64 Figure 15: McGuire Units 1 and 2 and Catawba Unit 1 Peak Cladding Temperature Elevation (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 46 of 64 Figure 16a: McGuire Units 1 and 2 and Catawba Unit 1 Vessel-Side Break Mass Flow Rate for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 47 of 64 Figure 16b: McGuire Units 1 and 2 and Catawba Unit 1 Pump-Side Break Mass Flow Rate for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 48 of 64 Figure 17: McGuire Units 1 and 2 and Catawba Unit 1 Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 49 of 64 Figure 18: McGuire Units 1 and 2 and Catawba Unit 1 Vapor Mass Flow Rate at the Top Cell Face of the Core Average Channel not Under Guide Tubes for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 50 of 64 Figure 19: McGuire Units 1 and 2 and Catawba Unit 1 RCS Pressure for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 51 of 64 Figure 20: McGuire Units 1 and 2 and Catawba Unit 1 Accumulator Injection Flow per Loop for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 52 of 64 Figure 21: McGuire Units 1 and 2 and Catawba Unit 1 Containment Pressure Comparison for the Region II Analysis
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 53 of 64 Figure 22: McGuire Units 1 and 2 and Catawba Unit 1 Vessel Fluid Mass for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 54 of 64 Figure 23: McGuire Units 1 and 2 and Catawba Unit 1 Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 55 of 64 Figure 24: McGuire Units 1 and 2 and Catawba Unit 1 Average Downcomer Collapsed Liquid Level (Relative to Bottom of Upper Tie Plate) for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 56 of 64 Figure 25: McGuire Units 1 and 2 and Catawba Unit 1 Safety Injection Flow per Loop (not including Accumulator Injection Flow) for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 57 of 64 Figure 26: McGuire Units 1 and 2 and Catawba Unit 1 Normalized Core Power Shapes for the Region II Analysis PCT Case Note: The localized power decreases occur at grid elevations.
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 of DPC-25-188, Revision 1 Page 58 of 64 Figure 27: McGuire Units 1 and 2 and Catawba Unit 1 Relative Core Power for the Region II Analysis PCT Case
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
- This record was final approved on 07/23/2025 21:17:00. (This statement was added by the PRIME system upon its validation)
U.S. Nuclear Regulatory Commission Serial: RA-25-0014 ENCLOSURE 5 10 CFR 50.46 EXEMPTION REQUEST FOR USE OF AXIOM FUEL CLADDING 3 PAGES PLUS THE COVER AXIOM, FULL SPECTRUM, FSLOCA, ZIRLO, and Optimized ZIRLO are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world.
All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
U.S. Nuclear Regulatory Commission Page 1 of 3 Serial: RA-25-0014 10 CFR 50.46 Exemption Request for Use of AXIOM Cladding 1.0 PURPOSE Pursuant to 10 CFR 50.12, Specific exemptions, Duke Energy Carolinas, LLC (Duke Energy),
requests an exemption from the requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, for Catawba Nuclear Station, Unit 1 (CNS U1) and McGuire Nuclear Station, Units 1 and 2 (MNS). This exemption request is related to the proposed use of the AXIOM fuel rod cladding material by CNS U1 and MNS since it does not conform to the specifications for either zircaloy or ZIRLO, both of which are explicitly identified as required fuel rod cladding material in 10 CFR 50.46, Section (a)(1)(i).
Consequently, an exemption is required from specific portions of 10 CFR 50.46 in order to support the application of AXIOM fuel rod cladding to the fuel for CNS U1 and MNS. The CNS U1 and MNS reloads scheduled to contain fuel rods with AXIOM cladding are proposed beginning Spring 2026 with CNS U1. This exemption request relates solely to the specific cladding material identified in these regulations (fuel rods with zircaloy or ZIRLO cladding) and will provide for the application of 10 CFR 50.46 to the use of AXIOM fuel rod cladding at CNS U1 and MNS.
2.0 BACKGROUND
As the nuclear industry pursues longer operating cycles, with increased fuel discharge burnup and fuel duty, the corrosion performance requirements for nuclear fuel cladding become more demanding. AXIOM cladding is designed to exhibit improved corrosion resistance, lower hydrogen pickup (HPU), and lower creep compared to other Westinghouse Electric Company, LLC (Westinghouse) cladding products (e.g., ZIRLO and Optimized ZIRLOTM). AXIOM cladding is a niobium-bearing alloy with reduced tin content to increase corrosion resistance like Optimized ZIRLO alloy. AXIOM cladding material has alloying elements including vanadium and copper to improve specific properties such as HPU.
In addition, fuel rod internal pressures (resulting from the increased fuel duty, use of integral fuel burnable absorbers, and corrosion/temperature feedback effects) have become more limiting with respect to fuel rod design criteria. Reducing the associated corrosion buildup, and thus, minimizing the temperature feedback effects, provides additional margin to the fuel rod internal pressure design limit.
As documented in the NRCs safety evaluation (SE) for Westinghouse topical report WCAP-18546-P-A, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, (ADAMS Accession No. ML22306A248), AXIOM cladding has been approved for use in PWR fuel. Additionally, an exemption request has previously been approved for use of AXIOM cladding by Turkey Point Nuclear Generating, Unit Nos. 3 and 4 (Turkey Point), by letter dated September 13, 2024 (ADAMS Accession No. ML24207A034).
Technical Specification (TS) changes for CNS U1 and MNS are required to allow the use of AXIOM fuel rod cladding for core reload applications. The request for these changes is provided in Enclosure 1 of this submittal.
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-25-0014 3.0 TECHNICAL JUSTIFICATION OF ACCEPTABILITY Westinghouse topical report WCAP-18546-P-A provides the details and results of tests for AXIOM cladding along with the material properties proposed for use in various models and methodologies when analyzing AXIOM fuel cladding, including the use of the FULL SPECTRUM' Loss-of-Coolant Accident (LOCA)(FSLOCA') evaluation model (EM) (ADAMS Accession No. ML17277A130). As described in the SE for WCAP-18546-P-A, the NRC staff reviewed the licensing criteria assessment, which included various fuel rod design criteria, safety analyses for both LOCA and non-LOCA transients, and radiological consequence analyses, and found that the topical report is acceptable for referencing in licensing applications to the extent specified under the limitations and conditions associated with the topical report.
The CNS U1 and MNS LOCA analysis for the fuel assemblies with AXIOM cladding was performed using the FSLOCA EM and adheres to the limitations of the associated topical reports. Enclosures 2 (proprietary) and 4 (non-proprietary) of the license amendment request attendant to this exemption request describes the CNS U1 and MNS LOCA evaluation performed for the fuel assemblies with AXIOM cladding.
4.0 JUSTIFICATION OF EXEMPTION 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not present an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security; and (4) special circumstances, as defined in 10 CFR 50.12(a)(2) are present. The requested exemption to allow the use of an advanced zirconium alloy other than zircaloy or ZIRLO for fuel cladding material, in this case AXIOM, at MNS and CNS U1 satisfies these criteria as described below.
- 1. This exemption is authorized by law As required by 10 CFR 50.12(a)(1), this requested exemption is authorized by law. The NRC has the authority under 10 CFR 50.12 to grant exemptions from the requirements of Part 50 upon showing proper justification. Further, it should be noted that CNS U1 and MNS are not seeking an exemption from the acceptance and analytical criteria of 10 CFR 50.46. The intent of this request is solely to allow the use of the criteria set forth in 10 CFR 50.46 for application to the AXIOM fuel rod cladding material, as it is currently not explicitly covered by the regulation.
- 2. This exemption will not present an undue risk to public health and safety The reload evaluations will ensure that acceptance criteria are met for future reload cores after the transition to fuel rods clad with AXIOM material. Fuel assemblies using AXIOM fuel rod cladding will be evaluated using NRC-approved analytical methods and plant-specific models to address the changes in the cladding material properties. The safety analyses for CNS U1 and MNS are supported by the applicable site-specific TSs. Reload cores are required to be operated in accordance with the operating limits specified in the TSs. Thus, the granting of this exemption request will not pose an undue risk to public health and safety.
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-25-0014
- 3. This exemption is consistent with common defense and security As noted above, this exemption request is only to allow the application of the aforementioned regulations to an improved fuel rod cladding material. All the requirements and acceptance criteria will be maintained. The special nuclear material in these assemblies is required to be handled and controlled in accordance with approved procedures. Use of AXIOM fuel rod cladding will not affect plant operations and is consistent with common defense and security.
- 4. Special circumstances are present which necessitate the request of an exemption to the regulations of 10 CFR 50.46 10 CFR 50.12(a)(2) states that the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemption meets the special circumstance criteria of 10 CFR 50.12(a)(2)(ii), which states: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. For MNS and CNS U1, application of the subject regulations is not necessary to achieve the underlying purpose of the rule.
10 CFR 50.46 identifies acceptance criteria for ECCS performance at nuclear power plants.
Westinghouse has performed an evaluation using LOCA methods as described in Enclosures 2 (proprietary) and 4 (non-proprietary) of this submittal to ensure that assemblies with AXIOM fuel rod cladding material meet all LOCA safety criteria.
5.0 CONCLUSION
The acceptance criteria and requirements of 10 CFR 50.46 are currently limited in applicability to the use of fuel rods with zircaloy or ZIRLO cladding. 10 CFR 50.46 does not apply to the proposed use of AXIOM fuel rod cladding material because AXIOM has a slightly different composition than zircaloy or ZIRLO. With the approval of this exemption request, these regulations will be applied to the use of AXIOM fuel rod cladding at CNS U1 and MNS.
In order to support the use of AXIOM fuel rod cladding material at CNS U1 and MNS, an exemption from the requirements of 10 CFR 50.46 is requested. The acceptance criteria applicable to AXIOM cladding have been established in WCAP-18546-P-A and the plant specific response is shown to comply with the acceptance criteria. Pursuant to 10 CFR 50.12, the requested exemption is authorized by law, does not present undue risk to public health and safety, and is consistent with the common defense and security. Approval of this exemption request does not violate the underlying purpose of the rule. In addition, special circumstances exist to justify the approval of an exemption from the subject requirements.