ML25212A131
| ML25212A131 | |
| Person / Time | |
|---|---|
| Site: | 99902041 |
| Issue date: | 08/07/2025 |
| From: | Ngola Otto Licensing Processes Branch |
| To: | Elliot G Framatome |
| Shared Package | |
| ML25212A128 | List: |
| References | |
| EPID L-2025-TOP-0013 | |
| Download: ML25212A131 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - CONTAINS PROPRIETARY INFORMATION SEPTEMBER 9-11, 2025, REGULATORY AUDIT PLAN FOR TOPICAL REPORT, BAW-10247, REVISION 0, SUPPLEMENT 3P, REVISION 0 REALISTIC THERMAL - MECHANICAL FUEL ROD METHODOLOGY FOR BOILING WATER REACTORS SUPPLEMENT 3: EXTENSION TO HIGHER EXPOSURES FRAMATOME, INC.
DOCKET NO. 99902041 EPID: L-2025-TOP-0013
1.0 BACKGROUND
By letter dated April 7, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25094A181), Framatome, Inc. (Framatome) submitted Topical Report (TR) BAW-10247, Revision 0, Supplement 3P, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology For Boiling Water Reactors Supplement 3: Extension to Higher Exposures (ADAMS Package Accession No. ML25094A177), to the U.S. Nuclear Regulatory Commission (NRC) for review and approval for licensing applications.
The NRC staff will perform a regulatory audit in accordance with Office of Nuclear Reactor Regulation 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 with discussions on technical issues, in determining whether requests for additional information are needed, and with drafting the safety evaluation.
2.0 REGULATORY AUDIT BASES Applicable regulations are found in the following sections of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities":
10 CFR Part 50, Appendix A, General Design Criteria 10, which require the reactor core, coolant, and protection systems are designed to provide sufficient margin to the specified acceptable fuel design limits (SAFDLs) during normal operation and including anticipated operational occurrences (AOOs).
Additional review guidance is provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 4.2, Fuel System Design. The regulatory audit will be held in accordance with the NRC procedures as described in LIC-111, Regulatory Audits.
3.0 REGULATORY AUDIT SCOPE The purpose of this audit is to review analyses and calculations described in the TR. An audit was determined to be the most efficient approach toward a timely resolution of questions associated with this review. The audit will provide efficiencies for the NRC staff by providing an opportunity for effective discussions with Framatome staff on the TR methodology and for clarifications on unresolved questions.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 4.0 LOGISTICS The NRC staff will conduct an in-person audit at Framatomes office in Richland for 2 days from September 9, 2025, to September 11, 2025, with an NRC staff member attending virtually. An electronic reading room (ERR) will be set up in advance to allow the NRC staff to review pertinent documentation. The ERR will be in accordance with the following restrictions:
The ERR will be password protected, and passwords will be distributed to NRC staff members directly involved in the BAW-10247, Supplement 3P, review on a need-to-know basis.
The ERR will not support any printing, saving, or downloading functions.
The NRC staff members given password access to the ERR will be informed of the conditions of use of the ERR by NRC Project Managers.
The conditions associated with the ERR will be maintained throughout its use during the BAW-10247, Supplement 3P, review process.
Documents posted to the ERR will be limited to basis and other reference material cited in the TR, unless otherwise agreed upon by Framatome and NRC management.
5.0 TEAM ASSIGNMENTS Key Framatome 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.
Kevin Heller, Technical Reviewer (NRR/DSS/SNFB)
Richard Fu, Technical Reviewer (NRR/DSS/SNFB) - virtual River Rohrman, Technical Reviewer (NRR/DSS/SNFB)
Ngola Otto, Project Manager (NRR/DORL/LLPB) 6.0 DISCUSSION TOPICS AND REQUESTED DOCUMENTS Discussion Topics The following general topics have been developed for discussion during the audit. Please have subject matter experts available for discussion of these general topics. More detailed questions can be found in Section 6 and may be developed during the remainder of the review.
Cladding Mechanical Properties for Both Stress-Relieved Annealed Zircaloy Cladding (SRA) and Recrystallized Annealed Zircaloy Cladding (RXA) Zircaloy-2 and -4 Uniform Zircaloy-2 BWR Corrosion/hydrogen pick-up (HPU) Model Dimensional change calculations related to rod growth and bowing Validation and Verification of the new HPU and oxidation model with the more recent data mentioned Reasoning during the development of the RODEX4 methodology update process and change thresholds for notification as it relates to the Limitation and Condition from the initial TR which requires NRC review to update values within the model
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Requested Documents The NRC staff requests the following documents to be made available via ERR 21 days in advance of the in-person audit to facilitate the staffs review and discussion in the regulatory audit.
Data and analysis that is associated with the mechanical tests on irradiated cladding of various metallurgical and zirconium alloy compositions to accompany Figures 3-9, Mechanical Testing on Irradiated Cladding, and 3-10, Burst Strains for Irradiated Cladding, Hydrogen < 500 PPM.
Data and analysis for the high and very high exposure fuel examined for rim width and high burnup structure (HBS) porosity in hot cells referenced in Section 4.2, High Burnup Structure.
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 safety evaluation 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:
General:
There does not appear to be discussion regarding the quantification of uncertainties associated with the new models presented for application to higher burnups or the existing models that are being extended to higher burnups. What is the quantification of associated uncertainties or the justification for the continued applicability of the existing uncertainty quantification?
Section 2.1, Applicable Regulatory Guidance
((
))
Section 2.4, Exposure Distributions in Part-length Fuel Rods Table 2-4, Data Gaps Identified in PNNL Report, provides the justifications for assessing ((
)) for approval of ((
)) for the following parameters for which significant data is only available up to 70 GWd/MTU: Fuel Centerline Temperature, Power Ramp Tests, Fission Gas Release, Cladding Corrosion, and Hydriding Cladding Mechanical Properties.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Section 3.0, RODEX4 Thermal-Mechanical Applicability to Extended Exposure Range Provide more details and data updates on extending Supplement 2 Safety Evaluation Report to allow modeling of BWR fuel assemblies from a fuel assembly-average exposure of
((
)) to ((
)).
Section 3.1.1, G1-Fuel Centerline Temperature The range of Cr-doped fuel temperatures tested does not include an extended burnup range.
Provide further explanation of how the Cr-doped fuel was treated for extended burnup past 62 MWD/kgU.
Section 3.1.2.2, High-Exposure FGR [Fission Gas Release] Dataset The discussions provided in Section 3.1.2.1, Summary of RODEX4 FGR Model, and Section 3.1.2.2, High-Exposure FGR Dataset, suggest that the database included in the original RODEX4 TR submittal for validation of the FGR model contains higher burnup data.
This database contains both steady state and transient (power ramp) data. However, its unclear from these discussions if the higher burnup portion of the data contained transient FGR. Has the FGR model been benchmarked or validated at extended burnup against the increased release of fission gases that can occur during rapid fuel pellet temperature redistribution that can occur during some transients?
Section 3.1.2.3, V&V of RODEX4 for the High-Exposure FGR
((
))
Section 3.1.2.4, Application of RODEX4 FGR Model to Cr-doped Fuel Provide an explanation to accompany Figure 3-6, Exposure Range for Cr-doped Fission Gas Release Dataset, as there are no references to it throughout the TR. ((
))
Section 3.2.1.1, Uniform Zircaloy-2 BWR Corrosion/HPU Model Would the approval of the new Uniform Zircaloy-2 BWR Corrosion/HPU model supersede the previous HPU model approved in BAW-10247, Supplement 1P? Are there any differences between the model in this TR versus Supplement 1P other than the new data that are considered? With the oxide layers no longer being overestimated due to a new measurement methodology that is not impacted by crud, will crud still be taken into consideration in a conservative manner?
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((
)) Please provide a comparison to data that demonstrates the predictive capability of the new model.
Section 3.2.1.2, V&V of the Uniform Corrosion Model for BWR Zircaloy-2 Cladding The full-length rod exposure for V&V of the Uniform Corrosion Model is stated to be ((
)) Please provide justification for the burnup values from plant A2 used for V&V for this new model.
Section 3.2.1.3, V&V of the HPU Model for BWR Zircaloy-2 Cladding
((
))
Section 3.2.2, G6-Cladding Mechanical Properties Please provide more information on the cladding mechanical properties and how they were determined. ((
))
Section 3.3.1, HP1-Oxidation and Hydriding Irradiation Limits The large margin to the HPU values provided are given based on an exposure of
((
)). Is this the full-length rod average or the fuel assembly average? The burnup range extension requested in this TR for a full-length rod average is ((
)). Please provide the hydrogen update margin at the peak of the burnup range requested or explain why what is in the TR is acceptable.
Section 3.3.3, HP3-Fission Gas Release and Fuel Rod Internal Pressure While the percentage FGR results of Section 3.1.2 provide support rod internal pressure is being calculated appropriately, a direct comparison of rod internal pressures against measured data makes for a ready assessment of the continued applicability of the current rod internal pressure limit at higher burnups (as well as assurance that all phenomenological models that contribute to rod internal pressure are being modeled appropriately). Are there any comparisons available of computed rod internal pressure versus measured rod internal pressure at higher burnups? If so, please provide the relevant analyses.
Section 4.2, High Burnup Structure There does not appear to be any reference to the correlation between rim width and HBS in the initial RODEX4 TR or safety evaluation. Provide more details on how this correlation was
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION developed, how RODEX4 implements it, and how it is acceptable at the burnup range requested in this TR.
Section 5.3, Extension of NAF RPP Correlations for Higher Enrichments Please provide more detail on how the higher enrichment data from GALILEO was incorporated in RODEX4.
TR Section 5.3, Extension of NAF RPP Correlations for Higher Enrichments, states that for AFM applications fuel enrichment must be increased above 5 weight-percent (wt%), and therefore, the RPP tables have been extended to 10 wt% U-235 enrichment and 10 wt%
gadolinia. However, it also states that laterafter the submittal of RODEX4 TRan updated versions of ((
)), namely ((
)) was used to extend the enrichment range to 10 wt% U-235, except for the gadolinia fuel, in which case the maximum U-235 enrichment remained at 5 wt% U-235. Please explain this apparent discrepancy in the range of uranium and gadolinia enrichments. Also, please provide detailed analyses of the extended enrichment range and gadolinia concentration.
Section 6.1.1, Rod Bow Since the rod bow correlation from Supplement 2P ((
)), how was the correlation extrapolated to the end-of-life burnup level requested? How is this different from the parameters that were assessed in Table 2-1, Standard Review Plan Section 4.2 Criteria, to
((
))? Is additional validation performed? If so, please provide a discussion and applicable data.
Section 7.0, Update Process Within the provided context, it is not clear how the word model is being used or how it should be interpreted, particularly with respect to replacing them. For example, changing the coefficients of an equation could be construed as a new model. Alternatively, changing coefficients could be considered the same model because the mathematical forms of the underlying equations remain unchanged, and only when the underlying mathematical forms of equations are changed would a model be considered new. In light of this, please further discuss the phrase the only change will be the replacement of the model or models in paragraph 3 of Section 7, specifically with respect to the intended interpretation of the word model.
Appendix A, Sample Problem
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Appendix B, BWR Fuel Rod to Fuel Assembly Differential Growth Correlation
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