RA-25-0020, License Amendment Request to Revise Reactor Core Safety Limit 2.1.1.2

License Amendment Request to Revise Reactor Core Safety Limit 2.1.1.2

ML25070A183
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	03/11/2025
From:	Gibby S Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-25-0020
Download: ML25070A183 (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 10 CFR 50.90 March 11, 2025 Serial: RA-25-0020 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 Revise Reactor Core Safety Limit 2.1.1.2 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, Units 1 and 2 (CNS) and McGuire Nuclear Station, Units 1 and 2 (MNS). The proposed amendment would revise the Reactor Core Safety Limit 2.1.1.2 for CNS and MNS to reflect the fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (ADAMS Accession No. ML17335A334).

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.

The enclosure to this license amendment request provides Duke Energys evaluation of the proposed changes. In addition, the attachment to the enclosure provides a copy of the existing TS pages marked with the proposed changes.

Approval of the proposed license amendment is requested within one year of acceptance. Once approved, the amendment will be implemented for CNS 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.

In accordance with 10 CFR 50.91, a copy of this application, with enclosure, is being provided to the designated North Carolina and South Carolina Officials.

This letter contains no regulatory commitments.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial: RA-25-0020 Please refer any questions regarding this submittal to Ryan Treadway, Director - Nuclear Fleet Licensing, at (980) 373-5873.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on March 11, 2025.

Sincerely, Shawn K. Gibby Vice President - Nuclear Engineering

Enclosure:

Evaluation of the Proposed Change

Attachment:

Proposed Technical Specification Changes (Mark-up) cc:

M. Miller, USNRC Region II - Regional Administrator (Acting)

D. Rivard, USNRC Resident Inspector - CNS C. Safouri, USNRC Senior Resident Inspector - MNS N. Jordan, USNRC NRR Project Manager - Fleet J. Minzer Bryant, USNRC NRR Project Manager - CNS J. Klos, 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-25-0020 Enclosure ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 8 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 8 Serial: RA-25-0020 Enclosure Evaluation of the Proposed Change License Amendment Request to Revise Reactor Core Safety Limit 2.1.1.2 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, Units 1 and 2 (CNS) and McGuire Nuclear Station, Units 1 and 2 (MNS). The proposed amendment would revise the Reactor Core Safety Limit (SL) 2.1.1.2 for CNS and MNS to reflect the fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (WCAP-17642-P-A, ADAMS Accession No. ML17335A334).

2.0 DETAILED DESCRIPTION

2.1 Background

System Design and Operation The sites are required by General Design Criteria (GDC) 10, Reactor design, to ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95%

probability at a 95% confidence level that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of CNS and MNS SL 2.1.1 prevent overheating of the fuel and cladding, as well as possible cladding perforation, which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the transient peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel.

Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Proper functioning of the Reactor Protection System (RPS) and the steam generator safety valves prevents violation of the reactor core SLs.

2.2 Current TS Requirements The existing TS SL 2.1.1.2 for CNS and MNS states:

U.S. Nuclear Regulatory Commission Page 2 of 8 Serial: RA-25-0020 Enclosure 2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080 degrees F, decreasing 58 degrees F for every 10,000 MWd/mtU of fuel burnup.

This SL is applicable in MODES 1 and 2.

2.3 Reason for the Proposed Change In order to ensure compliance with the SL, plant-specific safety analyses are performed. WCAP-17642-P-A, as reviewed and approved by the NRC, describes the fuel performance evaluation methodology and the PAD5 computer code, which is the principal design tool for evaluating fuel rod performance. This topical report defined the fuel pellet melting limit that is included within the PAD5 methodology based on available fuel pellet material properties. In addition to reviewing and approving the Westinghouse methodology, the NRC staff also concluded that the melting limits defined in WCAP-17642-P-A are acceptable. The proposed change to the CNS and MNS TS SL 2.1.1.2 will be implemented to maintain consistency between the value specified in the SL and the criteria used when performing confirmatory safety analyses that rely on the NRC-approved PAD5 methodology planned for implementation at CNS and MNS.

2.4 Description of the Proposed Change The proposed change revises CNS and MNS TS SL 2.1.1.2 to reflect the fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1. The revised version of SL 2.1.1.2 will read as follows:

2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080 degrees F, decreasing 9 degrees F for every 10,000 MWd/mtU of fuel burnup.

A mark-up of the proposed change to CNS and MNS TS Section 2.1.1 is provided in the attachment to this enclosure.

As part of the implementation actions associated with this change, the Updated Final Safety Analysis Reports (UFSARs) for CNS and MNS will be revised in accordance with 10 CFR 50.59 to address the updated fuel melt safety limit and its technical basis (e.g., CNS UFSAR Section 4.2.4.1.11 - Linear Heat Rate to melt, and MNS UFSAR Section 4.2.1.3.1.1.11 - Linear Heat Rate to Melt).

The current licensing basis safety analyses use the existing SL 2.1.1.2 for fuel melt as an acceptance criterion as required by the current methodology. Thus, CNS and MNS will continue to meet the existing SL when using their respective current licensing basis safety analyses, even with the implementation of the proposed SL. Since the existing SL for peak fuel centerline temperature is more restrictive than the proposed limit, the current licensing basis safety analyses will continue to meet the proposed SL following implementation of this license amendment request.

3.0 TECHNICAL EVALUATION

The Performance Analysis and Design (PAD) code is the principal design tool used by Westinghouse for evaluating fuel rod performance, where the computer program iteratively

U.S. Nuclear Regulatory Commission Page 3 of 8 Serial: RA-25-0020 Enclosure calculates the interrelated effects of fuel and cladding deformation including fuel densification, fuel swelling, fuel relocation, fuel rod temperatures, fill and fission gas release, and rod internal pressure as a function of time and linear power. PAD evaluates the power history of a fuel rod as a series of steady-state power levels with instantaneous jumps from one power level to another. The length of the fuel rod is divided into several axial segments, where each segment is assumed to operate at a constant set of conditions over its length. Fuel densification and swelling, cladding stresses and strains, temperatures, burnup and fission gas releases are calculated separately for each axial segment, and the effects are integrated to obtain the overall fission gas release and resulting internal pressure for each time step. The coolant temperature rise along the fuel rod is calculated based on the flow rate and axial power distribution. The cladding surface temperature is determined with consideration of corrosion effects and the possibility of local boiling.

Duke Energy currently self-performs fuel rod mechanical analyses for MNS and CNS with Westinghouses PAD4 code (WCAP-15063-P-A, Revision 1, ADAMS Accession No. ML003735452). The proposed revision to SL 2.1.1.2 will allow for the future transition to Westinghouses PAD5 code for fuel rod mechanical analyses. The PAD5 models are the latest versions of the Westinghouse PAD code, incorporating model updates that address all of the fuel and cladding performance models required for high burnup fuel design. Key fuel performance updates include the following: fuel thermal conductivity degradation (TCD) with burnup; enhanced high burnup athermal fission gas release (pellet rim effects); and enhanced high burnup fission gas bubble swelling. In order to reflect high burnup cladding performance, the cladding creep and growth models are also updated. In addition to high burnup analysis capability, a key driver for the implementation of the PAD5 models in fuel design is to address regulatory concerns associated with fuel TCD with burnup.

The proposed fuel centerline melt temperature within this license amendment request reflects the limit approved by the NRC for use within WCAP-17642-P-A. As stated in the Safety Evaluations (SEs) for the precedents listed in Section 4.2 of this request:

The empirically derived fuel centerline melt temperature described in WCAP-17642-P-A is based on fuel properties described in open literature. The description of the fuel properties can be found in (1) S.G. Popov; J.J. Carbajo; V.K. Ivanov; and G.L. Yoder,

'Thermophysical Properties of MOX and UO2 Fuels Including the Effects of Irradiation,"

ORNL/TM-2000/3S1 (2000) and (2) J.J. Carbajo; G.L. Yoder; S.G. Popov; and V.K.

Ivanov, "A Review of the Thermophysical Properties of MOX and UO2 Fuels," Journal of Nuclear Materials, 299, 181 (2001). As noted beneath the caption for Figure 59, on page 92 of the NRC staff SE on WCAP-17642-P-A (ADAMS Accession No. ML17257A338), in its approval of WCAP-17642-P-A, the NRC staff determined that this melting limit is acceptable. The burnup dependent fuel centerline melt temperature is based on inherent fuel properties and does not depend on any specific calculational methodology.

Therefore, the NRC staff considers it acceptable as a standalone limit.1 Where the footnote states:

1 The two identified references provide the data describing the fuel properties. The specific burnup dependence is provided in Section 6.1.5 of TR WCAP-17642-P-A based on an assessment of these data. The NRC staff determined that this burnup dependence was acceptable as described in Section 3.7.12, Pellet Overheating

U.S. Nuclear Regulatory Commission Page 4 of 8 Serial: RA-25-0020 Enclosure Melting, of the staffs SE approving TR WCAP-17642-P-A (ADAMS Accession No. ML17090A443).

In addition to updating the burnup-dependent term of the fuel melting limits in PAD5 based on journal-published fuel material data, additional validation performed in Section 2.1 of Appendix A of WCAP-17642-P-A shows that the PAD5 code, in conjunction with the new fuel melt limit, accurately predicts fuel melt based on comparisons to experimental observations. A comprehensive description of all PAD5 models, NRC Requests for Additional Information (RAI),

and the NRCs Safety Evaluation are also documented in WCAP-17642-P-A, Revision 1.

Additional discussion related to the Limitations and Conditions from the SE and how they will be satisfied for CNS and MNS is provided below in Section 3.1.

3.1 Limits of Applicability Duke Energys planned future implementation of PAD5 for CNS and MNS will comply with NRC Generic Letter (GL) 83-11, Supplement 1, Licensee Qualification for Performing Safety Analyses (ADAMS Accession No. ML031080345). Duke Energy has many years of experience performing fuel rod mechanical analyses (fuel melt, cladding strain, etc.) and currently self-performs fuel rod mechanical analyses for both Westinghouse and Framatome designed fuel.

Having self-performed these analyses since 1995 for Oconee Nuclear Station, Unit Nos. 1, 2, and 3 (ONS), 1999 for MNS and CNS, and 2019 for Sheaon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B. Robinson Steam Electric Plant, Unit 2 (RNP), Duke Energy uses the same software, methods, topical reports, and analysis guidelines that the vendors use for fuel mechanical analyses. Additionally, the vendors provide quality assurance for their software and update Duke Energy regarding change management on internal reports for software revisions and user manual updates. The vendors software is installed on Duke Energy servers and is controlled in accordance with Duke Energy software quality assurance procedures.

The proposed amendment will only be used in applicable safety analyses that are performed with the approved fuel performance methods in WCAP-17642-P-A. The pertinent Limitations and Conditions from the NRC SE for WCAP-17642-P-A that are applicable to the requested change to TS SL 2.1.1.2 are detailed below along with details regarding how each is satisfied for CNS and MNS.

Limitation and Condition (a): The NRC staff limits the applicability of the PAD5 code and methodology for cladding, fuel types and reactor parameters listed in Section 4.1 of the Safety Evaluation for WCAP-17642-P-A.

Response: The Westinghouse cladding and fuel utilized by CNS and MNS meet the constraints identified in Limitation and Condition (a). Furthermore, Duke Energy will apply PAD5 within the limits specified in Section 4.1 of the SE for WCAP-17642-P-A for cladding, fuel, and reactor parameters to be used at CNS and MNS. Since these PAD5 inputs depend on the reload design, these parameters are validated on a cycle-specific basis.

Limitation and Condition (b): The application of PAD5 should at no time exceed the fuel melting temperature as calculated by PAD5 due to the lack of properties for molten fuel in PAD5 and other properties such as thermal conductivity and fission gas release.

U.S. Nuclear Regulatory Commission Page 5 of 8 Serial: RA-25-0020 Enclosure Response: The peak fuel centerline temperature for CNS and MNS will be limited per this license amendment request. The requested amendment revises the peak fuel centerline temperature SL for CNS and MNS to ensure that fuel melt is precluded during conditions for normal operation and anticipated operational occurrences, consistent with GDC 10.

Regarding the remaining Limitations and Conditions associated with the PAD5 code, Limitation and Condition (c) reflects content that the NRC staff did not review since there was no mention of it in the revised TR. Therefore, it is not applicable to this license amendment request.

Additionally, Limitation and Condition (d) reflects that Westinghouse no longer requests approval of the model and methods improvement process (MMIP), and as such, the NRC does not approve the streamlined MMIP. This license amendment request is not pursuing approval of the MMIP, so this Limitation and Condition is also not applicable.

With respect to Limitation and Condition (e), the 10-year continued applicability review of PAD5 to ensure the best-estimate predictions and applied uncertainties remain valid is to be demonstrated and documented by Westinghouse, starting in 2017. As it relates to Duke Energy, Westinghouse provides quality assurance for the software and have an Appendix B quality assurance program that requires notification be provided to the utilities when there is a problem with the software. This will ensure Duke Energy is notified of any applicability issues for PAD5, with any identified issues to be addressed through the corrective action program.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Guidance 10 CFR 50.36 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.

Specifically, 10 CFR 50.36(c)(1) defines safety limits for nuclear reactors as the limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. The proposed amendment requests a change to the fuel centerline melt temperature safety limit.

10 CFR Part 50, Appendix A, General Design Criteria (GDC) 10 Appendix A, GDC 10, Reactor design, states:

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 operating, including the effects of anticipated operational occurrences.

The restrictions of SL 2.1.1.2 ensure that the requirements of GDC 10 are met in that it prevents overheating of the fuel and cladding by maintaining the steady state peak temperature below the level at which fuel centerline melting occurs. The proposed change

U.S. Nuclear Regulatory Commission Page 6 of 8 Serial: RA-25-0020 Enclosure will update the fuel centerline melt temperature limit in SL 2.1.1.2 to be consistent with the limit approved in WCAP-17642-P-A, Revision 1.

Conclusion The proposed change does not affect plant compliance with any of the above regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedents The following license amendments have been issued that revised the respective sites safety limits to reflect the fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1:

Millstone Power Station, Unit No. 3, Amendment No. 281 to Renewed Facility Operating License No. NPF-49 per letter dated January 7, 2022 (ADAMS Accession No. ML21326A099)

Surry Power Station, Units 1 and 2, Amendment Nos. 305 and 305 to Renewed Facility Operating License Nos. DPR-32 and DPR-37, respectively, per letter dated October 26, 2021 (ADAMS Accession No. ML21188A174).

Turkey Point Nuclear Generating Unit Nos. 3 and 4, Amendment Nos. 288 and 282 to Renewed Facility Operating License Nos. DPR-31 and DPR-41, respectively, per letter dated August 15, 2019 (ADAMS Accession No. ML19031C891).

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, Units 1 and 2 (CNS) and McGuire Nuclear Station, Units 1 and 2 (MNS). The proposed amendment would revise the Reactor Core Safety Limit 2.1.1.2 for CNS and MNS to reflect the fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (ADAMS Accession No. ML17335A334.).

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 change revises the Reactor Core Safety Limit 2.1.1.2 for CNS and MNS to reflect the fuel melt temperature specified in WCAP-17642-P-A, Revision 1, an NRC-reviewed and approved fuel performance code. The change does not require a physical change to plant systems, structures or components. Operations and analysis will continue to be in accordance with the licensing basis. The fuel melt temperature limit provides protection to the fuel and is consistent with the safety analysis. The proposed

U.S. Nuclear Regulatory Commission Page 7 of 8 Serial: RA-25-0020 Enclosure change does not impact either the initiation of an accident or the mitigation of its consequences. The proposed change does 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 change does 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 change revises a limit on fuel melt temperature in the CNS and MNS TS that is based on an NRC-reviewed and approved fuel performance code, and does not require physical changes to plant systems, structures, or components. Specifying a limit on the peak fuel centerline temperature ensures that the fuel design limits are met.

Operations and analysis will continue to be in compliance with NRC regulations.

Revising the fuel melt temperature limit does not affect any accident initiators that would create a new accident.

The proposed change neither installs or removes any plant equipment, nor alters the design, physical configuration, or mode of operation of any plant system, structure or component. The proposed change 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 accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. Specifically, no new hardware is being added to the plant as part of the proposed change, no existing equipment design or function is being modified, and no significant change in operations is being introduced. No new equipment performance burdens are imposed.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3)

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises a limit on fuel melt temperature in the CNS and MNS TS that is based on an NRC-reviewed and approved fuel performance code, and does not require physical changes to plant systems, structures, or components. Plant operations and analysis will continue to be in accordance with the licensing basis. Revising the fuel melt temperature limit as proposed will continue to ensure that applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their safety design functions and thereby margin of safety is not reduced.

U.S. Nuclear Regulatory Commission Page 8 of 8 Serial: RA-25-0020 Enclosure 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 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-25-0020 Enclosure - Attachment ATTACHMENT PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) 2 PAGES PLUS THE COVER

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