ML25205A085
| ML25205A085 | |
| Person / Time | |
|---|---|
| Site: | 07109356 |
| Issue date: | 09/15/2025 |
| From: | Storage and Transportation Licensing Branch |
| To: | NAC International |
| Shared Package | |
| ML25205A082 | List: |
| References | |
| EPID L-2025-LLA-0004 | |
| Download: ML25205A085 (1) | |
Text
Enclosure 2 SAFETY EVALUATION REPORT Docket No. 71-9356 Model No. MAGNATRAN Certificate of Compliance No. 71-9356 Revision 6
SUMMARY
By application dated December 19, 2024 (Agencywide Documents Access and Management System [ADAMS] Accession No. ML24355A131), as supplemented on June 26, 2025 (ML25177D088), NAC International (NAC or the applicant) applied for an amendment to Certificate of Compliance No. 9356 for the Model No. MAGNATRAN package. NAC requested an amendment to correct a licensing basis deficiency initially reported to the NRC on March 10, 2023 (ML23069A215). The report identified that a parameter used in the computation of bending stress in the finite element model used to structurally evaluate a fuel rod under the non-mechanistic tip-over accident condition was incorrectly specified resulting in the non-conservative calculation of stresses. MAGNATRAN safety analysis report (SAR) sections 2.11.1, 2.11.4, 2.11.6, and 2.12.1 have been revised to correct this error and provide additional clarification to the fuel rod evaluations. Also, by application dated August 13, 2025 (ML25225A114), NAC submitted a consolidated SAR.
Following staff review of the associated SAR, the staff finds that the changes do not affect the ability of the package to meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.
EVALUATION 2.0 STRUCTURAL EVALUATION The objective of the structural evaluation is to verify that the applicant has adequately evaluated the structural performance of the proposed transport package and demonstrated that it satisfies the regulations in 10 CFR Part 71, Packaging and Transportation of Radioactive Material.
The staff evaluated the proposed changes in the SAR, revision 25A, in accordance with the guidance of NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, US NRC, 2020 (ML20234A651). This section documents the staffs review, evaluation, and conclusions with respect to the structural safety aspects of the proposed transport package, as well as specific material property definitions.
2.1.
Description of Structural Design In this amendment, the applicant proposed the following design changes:
a.
the addition of the fuel rod evaluation for a 30-foot transport cask side drop event, b.
a revision to the fuel rod fatigue evaluation, and c.
a revision to the fuel rod evaluation for the 30-foot transport cask end drop event.
The following documents were submitted in support of this application amendment and were included with the application:
1.
71160-2139, PWR and BWR Fuel Assembly Fatigue Evaluation for MAGNATRAN, Revision 1.
2.
71160-2140, Fuel Rod Evaluation for the MAGNATRAN 30-ft Side Drop Accident, Revision 1.
2 3.
71160-2126, Fuel Assembly Structural Evaluation for the MAGNATRAN End Drop Condition, Revision 7.
2.1.1 Fuel Rod Evaluation for 30-foot Side Drop The applicant documented the analyses of fuel rods for the Hypothetical Accident Condition (HAC) 30-foot transport cask side drop in calculation 71160-2140 and described it in SAR section 2.11.4. The applicant performed a number of evaluations to represent the effects of this HAC event on 33 PWR and 27 BWR high burnup fuel assemblies, each comprised of 7 fuel types, at a bounding fuel temperature of 400 °C (752 °F), the maximum permitted cladding temperature during normal transport conditions. The applicant employed the finite element program ANSYS for these analyses.
Each evaluation considered a single, empty fuel rod with the cladding weight adjusted for the mass of the missing fuel and applied a cladding flexural rigidity factor of 1.25 to adjust for the absence of the fuel pellet, as recommended by section 2.3.4 of NUREG-2224, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel, USNRC, November 2020 (ML21091A321).
Cladding thicknesses for both PWR and BWR rods are also reduced to account for oxide layers for the various cladding materials, as recommended by figure 2-5 of NUREG-2224. For the PWR fuel rod analyses, the applicant considers PWR fuel rods with and without grid damage.
For the PWR fuel rod without grid damage, the applicant considers the nominal as-built spacing between grids, while for the PWR fuel rod with grid damage, a maximum unsupported length of 60 inches is considered to account for fuel rods with missing, slipped, or damaged grids, which the applicant determines to be bounding over a configuration with no missing/damaged grids.
For the BWR fuel rod analyses, the applicant considers the unsupported lengths to be the nominal, as-built configuration. For both PWR and BWR fuel rods, the applicant limits deflections in the analyses by the space between adjacent fuel rods as well as that between the fuel tube wall. The staff finds the applicants choice of analysis parameters and method of evaluation to be acceptable as they are either conservative or follow the guidance provided.
For the PWR rod analyses, the applicant considered high burnup conditions and cladding alloys that are typical for PWR fuel. Similarly, the applicant considered high burnup conditions and cladding alloys that are typically used for BWR fuel. The fuel rod and fuel pellet material properties of elastic modulus, density and yield strength, as applicable, are taken from various references.
The staff reviewed the applicants selection of material properties for the cladding alloys to support the analyses. In addition, the staff reviewed information on cladding material properties not cited by the applicant including:
Shimskey, R., et al. FY2014 PNNL Zr Cladding Testing Status Report, PNNL-23594, August 30, 2014.
Wells, B.E., et al. Evaluation of Increased Peak Temperatures for Spent Fuel Cladding Performance during Dry Storage, PNNL-30430, Rev. 1, September 2020.
The staff determined that the material properties for the BWR fuel cladding alloys used by the applicant in their side drop evaluation were acceptable because the values were obtained by measurements on irradiated cladding alloys at temperatures consistent with the applicants analyses. Similarly, the staff determined that the material properties for most of the PWR fuel cladding alloys used by the applicant were obtained by measurements on the irradiated cladding samples at elevated temperatures and were therefore acceptable.
3 The staff noted that the data used by the applicant for one of the PWR cladding alloys was obtained using tensile tests under non-quasi-static testing conditions. Zirconium-based fuel cladding alloys are known to strain harden as a function of strain rate which, in turn, increases the measured yield strength. Based on the information provided in Shimskey et al. (2014) and Wells et al., (2020), the staff determined that the yield strength of the cladding alloy would be increased by approximately 6 percent under the testing conditions in the reference cited by the applicant. The staff also determined that publicly available data for irradiated cladding material properties at elevated temperatures is limited and alternative references for properties for the alloy under quasi-static testing conditions are not available. After reviewing the available data and the applicants analysis, the staff determined that the cited material properties for the cladding alloy were acceptable because: (1) the measurements were conducted using irradiated materials over a range of temperatures that bound the applicants analyzed maximum temperature for the cladding alloy under transportation conditions, (2) the strain rates under drop accident conditions would be greater than quasi-static strain rates typically used to determine material properties, and (3) use of yield strength as an acceptance criteria is conservative because all zirconium-based cladding alloys strain harden above the yield stress and retain measurable ductility after irradiation.
The acceleration values applied to each fuel rod in the analyses were determined through the use of a dynamic load factor based on the modal analysis of each fuel rod, following the general principles presented in section 2.3.5.2 of NUREG-2224. The applicant created a broadened and enveloped response spectra from the acceleration time histories of the LS-DYNA finite element analyses of the 30-foot cask drops. The resulting acceleration values, applied uniformly along the entire length of the fuel rods in the analyses, were 28 g for the damaged PWR rods with grid damage, 48 g for the PWR rods without grid damage, and 48 g for the BWR rods. The staff finds this method of acceleration determination to be acceptable as it follows guidance and the resulting applied acceleration values to be acceptable as they are conservative.
For the final analysis, one PWR fuel rod with grid damage and one BWR fuel rod are chosen by the applicant based on the associated cladding section properties and inertia loading that are expected to produce bounding cladding stresses during the 30-foot transport cask side drop event. For both the PWR and BWR rod analyses, the factor of safety for all cladding yield stress versus actual stresses were greater than a value of one for all material types, indicating that the fuel rod cladding material meets the acceptance criteria during the HAC 30-foot cask side drop event. Based on these results, the staff finds the fuel rods to be structurally adequate for the HAC event required by 10 CFR 71.73(c)(1).
2.1.2 Fuel Rod Evaluation for 30-foot End Drop The applicant documented the reduced factors of safety for the HAC 30-foot transport cask end drop fuel rod analyses in revisions to calculation 71160-2126 and tables in SAR sections 2.11.1 and 2.11.1.2 resulting from the reduced M5 fuel clad yield strengths at 400 °C (752 °F), the maximum permitted cladding temperature during normal transport conditions. The applicant previously determined that the PWR fuel rod analyses, both intact and damaged grids, bounded those of the BWR rod analyses, and the M5 cladding material also bounded the Zirlo and Zirc-4 materials. The resulting revised factors of safety for the PWR cladding yield stress versus actual stress were greater than one for the M5 material type with or without grid damage, indicating that the fuel rod cladding material meets the acceptance criteria during the HAC 30-foot cask end drop event. Based on these results, the staff finds the fuel rods to be structurally adequate for the HAC event required by 10 CFR 71.73(c)(1).
4 2.1.3 Fatigue Evaluation of Fuel Rods The applicant documented this Normal Conditions of Transport (NCT) analysis in calculation 71160-2139 and revised SAR section 2.11.6 to describe this evaluation. The applicant performed a number of evaluations to represent the effects of this NCT on 33 PWR and 27 BWR high burnup fuel assemblies, each comprised of 7 fuel types, at a bounding fuel temperature of 400 °C (752 °F), the maximum permitted cladding temperature during normal transport conditions. The applicant employed the finite element program ANSYS for these analyses.
Each evaluation considered a single, empty fuel rod with the cladding weight adjusted for the mass of the missing fuel and applied a cladding flexural rigidity factor of 1.25 to adjust for the absence of the fuel pellet, as recommended by section 2.3.4 of NUREG-2224. The applicant conservatively reduces the cladding thicknesses for the PWR rods by 125 microns and by 120 microns for the BWR rods to account for oxide layers, as recommended by figure 2-5 of NUREG-2224. For the PWR fuel rod analyses, the applicant considers PWR fuel rods with and without grid damage. For the PWR fuel rod without grid damage, the applicant considers the nominal as-built spacing between grids, while for the PWR fuel rod with grid damage, a maximum unsupported length of 60 inches is considered to account for fuel rods with missing, slipped, or damaged grids, which the applicant determines to be bounding over a configuration with no missing/damaged grids. For the BWR fuel rod analyses, the applicant considers the unsupported lengths to be the nominal, as-built configuration. Both PWR and BWR fuel rods, deflections are limited in the analyses by the space between adjacent fuel rods as well as that between the fuel tube wall. The staff finds the applicants choice of analysis parameters and method of evaluation to be acceptable as they are either conservative or follow the guidance provided.
The applicant performed response spectra analyses for the fuel rods employing five test cases from the ENSA/DOE rail cask test documented in SAND2018-13268R, Data Analysis of ENSA/DOE Rail Cask Tests, Spent Fuel and Waste Disposition, US Department of Energy, Spent Fuel and Waste Science and Technology, November, 2018, which include acceleration data in three orthogonal directions up to a frequency of 1,000 Hz. The applicant reported the resulting maximum fuel rod cladding stress and strain for the analyzed rods and compares the strain values to table 2-6 and figure 2-12 of NUREG-2224, both of which indicate lower-bound fatigue limits for high burnup Zirc-4 fuel cladding. For both the PWR and BWR rod analyses, the resulting strain values were lower than the 0.060% limits for cycles beyond 2.74 x 105 in NUREG-2224, indicating that the fuel rod cladding material meets the acceptance criteria during fatigue conditions under NCT. Based on these results, staff finds the fuel rods to be structurally adequate for the NCT event required by 10 CFR 71.71(c)(5).
2.2.
Evaluation Findings The staff reviewed the amendment package for the revised NCT and additional HAC analyses and concludes that it satisfies the requirements of 10 CFR 71.71(c)(5) and 71.73(c)(1).
The staff reviewed the structural and material performance of the package under the NCT required by 10 CFR 71.71 and the HAC required by 10 CFR Part 71.73 and concludes that it satisfies the requirements of 10 CFR 71.51(a)(1) and 71.51(a)(2) for a Type B package and 10 CFR 71.55(d)(2) and 71.55(e) for a fissile package.
Based on review of the statements and representations in the application amendment request, the NRC staff concludes that the MAGNATRAN package has been adequately described and
5 evaluated to demonstrate that it satisfies the structural integrity and material performance requirements of 10 CFR Part 71.
CONDITIONS In addition to small editorial changes, the following changes have been made to the certificate:
Condition 3(b) has been revised to reflect the date of the consolidated application.
Condition No. 14 has been edited to read Revision No. 5 of this certificate may be used until August 31, 2026.
The references section has been updated to reflect the date of the consolidated application.
CONCLUSION Based on the statements and representations in the application, as supplemented, and the conditions listed above, the staff concludes that the Model No. MAGNATRAN package design has been adequately described and evaluated, and that these changes do not affect the ability of the package to meet the requirements of 10 CFR Part 71.
Issued with Certificate of Compliance No. 9356, Revision No. 6.