ML25325A081
| ML25325A081 | |
| Person / Time | |
|---|---|
| Site: | |
| Issue date: | 11/25/2025 |
| From: | Lenning K Licensing Processes Branch |
| To: | |
| References | |
| EPID L 2025 TOP 0020 | |
| Download: ML25325A081 (0) | |
Text
Enclosure 1 U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION AUDIT PLAN FOR THE REGULATORY AUDIT OF WESTINGHOUSE ELECTRIC COMPANY TOPICAL REPORT WCAP-18951-P/NP, REVISION 1, TRITON11 REFERENCE FUEL DESIGN WESTINGHOUSE ELECTRIC COMPANY DOCKET NO. 99902038 EPID: L-2025-TOP-0020
1.0 BACKGROUND
By letter dated May 28, 2025 (Agencywide Documents Access and Management System Accession No. ML25149A002), Westinghouse Electric Company (Westinghouse) submitted Topical Report (TR) WCAP-18951-P/NP, Revision 1, TRITON11 Reference Fuel Design (Proprietary/Non-Proprietary), for the U.S. Nuclear Regulatory Commission (NRC) staffs review and approval. This TR describes improvements to the previously approved boiling water reactor (BWR) fuel mechanical design methodology intended to support fuel design and licensing applications up to [
] This TR also provides a reference product description, including mechanical specifications and performance aspects, of the TRITON 11 fuel assembly design. The primary focus of the audit is to review analyses and calculations described in the TR.
The NRC staff plans to conduct a regulatory audit on December 9-10, 2025, in accordance with Office of Nuclear Reactor Regulation (NRR) Office Instructions LIC-111, Revision 2, Regulatory Audits (ML24309A281), and LIC-500, Revision 9, Topical Report Review Process (ML20247G279). The audit will facilitate the NRC staff discussions on technical issues, in determining whether requests for additional information are needed and by drafting the safety evaluation (SE).
2.0 REGULATORY AUDIT BASES Regulatory guidance for the review of fuel system designs and adherence to applicable General Design Criteria (GDCs) is provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP) Section 4.2, Fuel System Design. As stated in Section 4.2 of SRP:
The fuel system safety review provides assurance that (1) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs), (2) fuel system damage is never so severe as to prevent control rod
insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained.
GDC 10, within Appendix A to 10 CFR Part 50, also addresses item 1 above. Specifically, GDC 10 establishes specified acceptable fuel design limits (SAFDLs) that should not be exceeded during any condition of normal operation, including the effects of AOOs. Therefore, the SAFDLs are established to ensure that the fuel is not damaged. Within this context, not damaged means that the fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analysis. The design limits of GDC 10 (i.e., the SAFDLs) accomplish these objectives. In a fuel rod failure, the fuel rod leaks, and the first fission product barrier (the cladding) is breached.
The dose analysis required by 10 CFR Part 100 for postulated accidents must account for fuel rod failures. Coolability, in general, means that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat even after a severe accident. The general requirements to maintain control rod insertability and core coolability appear repeatedly in the GDC found in Appendix A to 10 CFR Part 50 (e.g., GDC 27 and 35). In particular, 10 CFR 50.46 provides the specific coolability requirements for the loss-of-coolant accident (LOCA).
The NRC staffs review of WCAP-18951-P/NP, Revision 1, is to ensure that the mechanical design methodology adequately addresses the applicable regulatory requirements identified in SRP Section 4.2. In addition, the NRC staff reviewed the TRITON 11 fuel assembly design to ensure its performance satisfies these requirements. The regulatory audit will be held in accordance with the NRC procedures as described in LIC-111, Regulatory Audits.
3.0 LOGISTICS The audit will be started once an electronic reference portal is set up and the documentation is made available to the NRC staff. The initial desk audit will be conducted over several weeks.
Westinghouse will be kept informed on a regular basis during periodic discussions with the project manager regarding the progress.
The audit plan was issued on September 11, 2025, for the hybrid audit at NRC Headquarters in Rockville, MD, originally scheduled for October 14-15, 2025 (ML25248A337). The audit plan was revised to reflect new audit dates and location. No other changes to the audit plan were necessary. The NRC staff will perform a hybrid (in-person at Westinghouse facility in Cranberry, Pennsylvania, and virtually) regulatory audit on December 9-10, 2025.
4.0 AUDIT TEAM Key Westinghouse personnel involved in the development of the TR should be made available for interactions on a mutually agreeable schedule to respond to any questions from the NRC staff.
Team Member Division Area of Responsibility Kate Lenning NRR/DORL/LLPB Project Management Richard Fu NRR/DSS/SFNB Technical Review Jeremy Dean NRR/DSS/SFNB Technical Review
5.0 DISCUSSION TOPICS AND REQUESTED DOCUMENTS Discussion Topics Detailed questions can be found in Section 8.0 and may be further developed during the remainder of the review.
Requested Documents The NRC staff requests the following documents to be made available via Westinghouse electronic portal in advance of the in-person audit to facilitate the NRC staffs review and discussion in the regulatory audit.
Data and analysis associated with Figure 4-2 Geometrical Compatibility TRITON11 beginning-of-life (BOL)/non-TRITON11 Fuel BOL Data and analysis associated with Figure 4-6 Predicted and Measured end-of-life (EOL)
Channel Creep Deformation (Change in Channel Width) at Thick/Thin Channel Section Data and analysis associated with Figure 4-7 Predicted and Measured EOL Channel Creep Deformation (Change in Channel Width) at One (Thin) Thickness Channel Section Data and analysis associated with Figure 4-21 Total Hydrogen Concentration versus Burnup Data and analysis associated with Figure 4-23 Rod Average Oxide Thickness Data and analysis associated with Figure 4-24 Rod Maximum Oxide Thickness Data and analysis associated with Figure 4-26 Example of Transient Power History (AOO) for Maximum Temperatures 6.0 SPECIAL REQUESTS The NRC staff would like access to the documents listed above in Section 5.0 through an online portal that allows the NRC staff to access documents via the internet. The following conditions associated with the online portal must be maintained throughout the duration that the NRC staff have access to the online portal:
The online portal will be password-protected, and separate passwords will be assigned to the NRC staff who are participating in the audit.
The online portal will be sufficiently secure to prevent the NRC staff from printing, saving, downloading, or collecting any information on the online portal.
Conditions of use of the online portal will be displayed on the login screen and will require acknowledgement by each user.
Username and password information should be provided directly to the NRC staff. The NRC project manager will provide Westinghouse with the names and contact information of the NRC staff who will be participating in the audit. All other communications should be coordinated through the NRC project manager.
7.0 DELIVERABLES The NRC team will develop an audit summary report to convey the results. The report will be placed in ADAMS within 90 days of the completion of the final audit session. The audit information the NRC staff determines to be necessary to support the development of the NRC staffs SE will be requested to be submitted on the docket.
8.0 AUDIT QUESTIONS AND REQUESTS The NRC staff have identified questions and information requests in advance to assist the preparation of materials and subject matter experts available for the audit. Noting the early stages of review, the NRC staff may develop more questions during the remainder of the review.
The questions and information requests identified in advance are as follows:
Section 2.1 Assembly Description Please discuss the function of the balance plate and how it achieves that function. Does this create any new stress/strain forces on the water rods while handling or during operation?
Section 2.3 Fuel Channel Please discuss any differences in channel growth, bulge, or bow as it relates to the two channel variants offered.
Section 2.4 Plant Dependent Features Please discuss the ranges of the options for plant specific modifications to the following and their impacts on the plants licensing analyses:
[
]
Section 4.2 Fuel Assembly Components Evaluation What EOL equivalent channel burnup is expected and are there any related burnup limits associated with the fuel channel? Does any data exist [ ] Megawatt-days per kilogram of uranium (MWd/kgU) that would demonstrate no breakaway channel growth behavior is expected in this region (Figure 4-1 in the TR)?
Does any data exist [ ] MWd/kgU that would demonstrate no breakaway channel bulge behavior is expected in this region (Figure 4-8 in the TR)?
What are your TRITON11 predictions of the channel creep deformation for assembly burnup at
[ ] MWd/kgU based on Figure 4-6 and Figure 4-7 in the TR?
Does any data exist [ ] MWd/kgU that would demonstrate that fuel rod growth behavior is expected in this region (Figures 4-9 and 4-10 in the TR)?
Are there any oxide thickness data at [ ] MWd/kgU assembly burnup for TRITON11 based on Figure 4-11 and Figure 4-12 in the TR?
Section 4.3 Fuel Rods Evaluation Please describe the process for applying a specific treatment on uncertainties methodology (Monte Carlo, root-mean-square, or bounding/deterministic) to any fuel rod evaluation of how that process is captured in a plants licensing basis. Does the treatment change from cycle to cycle and, if so, are there any restrictions for changing the uncertainty treatment?
What are the EOL limits associated with the fuel rod hydrogen content? Does any data exist
[ ] MWd/kgU that would demonstrate fuel rod hydrogen behavior is expected in this region (Figures 4-21 and 4-22)?
Does any oxide thickness data exist for fuel rod average burnup [ ] MWd/kgU (PLR) that would demonstrate fuel rod cladding corrosion behavior of HiFi and TRITON11 is expected in this region (Figures 4-21 and 4-22)?
Section 7.0 Operating Experience Based on the data in Table 7-2, Operating Experience and Post Irradiation Examination (PIE) of TRITON11 components and materials by January 2025, Section 7.4.2 of the TR, please explanation in detail how the data support TRITON11 licensing to [ ] MWd/kgU rod average.
Section 9.0 Testing, Inspection, and Surveillance Plans This TR is based on Nordic lead test assemblies operations under noble water chemistry conditions. How does this data apply to use of TRITON11 with typical U.S. plants conditions?
Section Appendix A Applicability of Boiling Water Reactor Methods for ADOPT' Fuel and Extended Burnup Range The NRC staff will develop audit questions for Appendix A after completion of the first audit, and, if necessary, schedule a separate audit for Appendix A later.