ML23193A976
| ML23193A976 | |
| Person / Time | |
|---|---|
| Site: | 07109235 |
| Issue date: | 07/27/2023 |
| From: | Storage and Transportation Licensing Branch |
| To: | NAC International |
| Shared Package | |
| ML23164A263 | List: |
| References | |
| CoC No. 9235, EPID L-2022-LLA-0070 | |
| Download: ML23193A976 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Model No. STC Docket No. 71-9235 Certificate of Compliance No. 9235 Revision 24
SUMMARY
By letter dated May 9, 2022 (Agencywide Documents Access and Management System
[ADAMS] Accession No. ML22130A773), as supplemented on April 10, 2023 (ADAMS Accession No. ML23100A071), NAC International submitted an application to revise Certificate of Compliance (CoC) No. 9235 for the Model No. NAC-STC package. The applicant requested to credit the flexural rigidity of the fuel pellet as recommended in NUREG-2224 versus the previously approved licensing approach for directly loaded high burnup (HBU) fuel of time limitations associated with ductile to brittle transient temperature limits as well as revise HBU fuel discussions in safety analysis report (SAR) chapters 2 and 3. The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the application, including its supplement, using the guidance in NUREG-2216, Standard Review Plan for Spent Fuel and Transportation Packages for Radioactive Material. Based on the statements and representations in the application, as supplemented, and the conditions listed below, the staff concludes that the packages meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of Radioactive Material.
1.0 GENERAL INFORMATION 1.1 Requested Changes On August 29, 2014, the applicant submitted revision 14B for the Model No. NAC STC package to authorize the shipment of HBU fuel. Revision 14B added sections 1.1.1 and 1.1.2 to the SAR General Information chapter. These sections described the applicants licensing approach for authorizing the shipment of HBU fuel using ductile to brittle transient temperature data. The applicant chose to delete these SAR sections and take credit for the flexural rigidity of HBU fuel rod pellets in the spent fuel content of the NAC-STC following the recommendations provided in NUREG-2224, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material. Based on the findings in safety evaluation report (SER) sections 2, 3 and 7, the staff finds deleting these sections acceptable.
1.2 Evaluation Findings
Based on a review of the statements and representations provided by the applicant, the staff concludes that the contents have been adequately described to meet the requirements of 10 CFR Part 71.
Enclosure 1
2 2.0 STRUCTURAL EVALUATION The staff reviewed and evaluated the changes proposed by the applicant in the SAR, revision 22A. The specific proposed changes for evaluations in this section included:
- Addition of a new structural analysis to the current SAR Section 2.13.6.15.2, Fuel Rod Assessment for HBU Fuel for 30-foot Side Drop.
- Addition of a new SAR Section 2.13.6.15.4, Fatigue Evaluation for HBU Fuel for Normal Conditions of Transport.
This SER section documents the staffs reviews, evaluations, and conclusions with respect to structural safety of the fuel rods.
2.1 Evaluation for Addition of Structural Analysis to the SAR Section 2.13.6.15.2, Fuel Rod Assessment for HBU Fuel for 30-foot Side Drop The applicant performed an additional structural analysis to calculate stresses in pressurized water reactor (PWR) HBU fuel rods (WE 17x17) during a cask side drop under hypothetical accident conditions (HAC) with the ANSYS finite element (FE) program. Using a bounding acceleration of [ ], the applicant calculated the stresses using the PWR fuel rod model in revision 18 of the SAR (reference 2.1) which the staff previously reviewed and accepted. The applicant followed the methodology in section 2.3 of NUREG-2224 (Reference 2.3) to calculate the HBU fuel rod cladding stresses.
The SAR appendix D provided the maximum bending stress for the PWR HBU fuel rod (WE17x17) analytical results calculated using the ANSYS FE program. The applicant noted that the maximum bending stress of [ ] ksi previously presented in rev. 20 of SAR section 2.13.6.15.2, (reference 2.2) was erroneous due to an incorrect input computer file for the model generation. However, the new maximum bending stress of [ ] ksi remained below the minimum yield strength of the fuel rod clad of [ ] which is the maximum fuel cladding temperature specified in revision 20 of the SAR for the normal conditions of transport (NCT). This resulted in a factor of safety, a ratio of the maximum calculated stress with respect to the allowable stress, greater than one. Based on the results of the structural analysis, the applicant concluded that the PWR HBU fuel rods remain in an elastic range and are structurally adequate for a side drop accident under HAC.
The staff reviewed the applicants structural analysis. The staff found that the results of the analysis showed that the PWR HBU fuel rods remain structurally adequate for a side drop accident under HAC. Based on its reviews, the staff finds that the applicants additional structural analysis in SAR section 2.13.6.15.2, Fuel Rod Assessment for HBU Fuel for 30-foot Side Drop, is acceptable because (i) the staff previously reviewed and accepted the FE model for the HBU fuel rods, (ii) the applicant followed the methodology suggested in NUREG-2224, and (iii) the calculated maximum bending stress from the structural analysis is less than the yield stress which indicates that the PWR HBU fuel rods are structurally adequate for a side drop accident under HAC because they remain in an elastic range.
3 2.2 Evaluation for Addition of a new Section 2.13.6.15.4, Fatigue Evaluation for HBU Fuel for Normal Conditions of Transport The applicant performed a fatigue evaluation of PWR HBU fuel rods (WE 17x17) under NCT for the NAC-STC system. The applicant developed a FE model to calculate the stress and strain in the fuel cladding with the ANSYS FE program. The model represented the fuel cladding using the ANSYS 3-D BEAM4 element. The applicant adjusted the clad density to account for the mass of the fuel pellet and modeled the grid locations as simple supports in the lateral directions. The ANSYS program performed the spectrum analyses response for the fuel rods using the transport cask platform response spectra from seven test cases as documented in the ENSA/DOE rail cask test (reference 2.4).
The applicant presented the proprietary analytical results in a table on SAR page 2.13.6-59.
The table identified the maximum PWR fuel rod stress and strain for the seven cases. The maximum strain of [ ] among the seven cases proved to be well below the 0.06% end point of the Lower-Bound Fatigue Curve shown in table 2-5 and figure 2-12 of NUREG-2224 (reference 2.3). The applicant stated that fatigue of the PWR fuel rod is not a concern for the HBU PWR fuel assemblies for transport conditions because the calculated maximum strain of [
] is well below the 0.06% end point of the Lower-Bound Fatigue Curve as shown in table 2-5 and figure 2-12 of NUREG-2224 (reference 2.3). Therefore, the applicant stated that PWR fuel rod fatigue is not a concern for HBU PWR fuel assemblies for transport conditions.
However, the staff noticed that the calculated stress and strain values were somewhat peculiar when compared to each other. As a result, the staff issued a request for additional information regarding the maximum rod strain calculation. The applicant responded that there was an error in the computer input file, and it reperformed the analyses after correcting the input file error.
The applicant stated that: (i) the input error was limited to the fuel rod beam element stress and strain reported, (ii) there was no impact on either the stiffness matrix or the mass of the rod, (iii) the identical methodology was used for the updated analyses, (iv) the tables in section 2.13.6.15.4 of the SAR and NAC Calculation No. 423-2020 were revised as shown in the table below, and (v) PWR fuel rod fatigue is still not a concern for the HBU PWR fuel assemblies under transport conditions because the calculated new maximum strain is below the 0.06% end point of the Lower-Bound Fatigue Curve shown in table 2-5 and figure 2-12 of NUREG-2224 (reference 2.3). The staff reviewed the applicants statement and confirmed that the calculated maximum fuel rod strain is less than the maximum strain of 0.06% from the fatigue curve of the model tests in NUREG-2224.
Case No. Max Stress (ksi) Max Strain (%)
1 [ ] [ ]
2 [ ] [ ]
3 [ ] [ ]
4 [ ] [ ]
5 [ ] [ ]
6 [ ] [ ]
7 [ ] [ ]
The staff reviewed the evaluations and concluded that fatigue is not a concern of the HBU PWR fuel assemblies during a transport under NCT because (i) the FE model representing a single
4 fuel rod was adequately developed for the ANSYS FE structural analyses, and (ii) the calculated maximum strain is less than the maximum strain of 0.06% from the fatigue curve of the model tests in NUREG-2224.
2.3 Evaluation Findings
The staff reviewed the applicants evaluations for the PWR HBU fuel rods during a side drop under HAC and a transport under NCT. The staff concludes that the results of the evaluations are acceptable, and that the HBU fuel rods remain structurally adequate for the side drop accident under HAC and that fatigue of the HBU fuel rods is not a concern during a transport under NCT. The staff finds that the PWR HBU fuel rods have adequate structural integrity to meet the structural requirements of 10 CFR Part 71.
2.4 References 2.1 NAC International, Safety Analysis Report (SAR) for the NAC Storage Transport Cask (NAC-STC), Revision 18, March 2017.
2.2 NAC International, Safety Analysis Report (SAR) for the NAC Storage Transport Cask (NAC-STC), Docket No. 71-9235, Revision 20, July 2019.
2.3 NUREG-2224, Dry Storage and Transportation of HBU Spent Nuclear Fuel - Final Report, November 2020.
2.4 SAND2018-13258R, 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.
3.0 THERMAL EVALUATION The applicant applied for NRC review of a revision to CoC No. 9235, (revision 23) to specifically credit the flexural rigidity of the fuel pellets in HBU fuel rods that make up spent fuel contents of the NAC-STC following the recommendations provided in NUREG-2224, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material.
The applicant used the analysis in SAR section 2.13.6.15.4, Fatigue Evaluation for HBU Fuel for Normal Conditions of Transport to justify removal of the licensing approach for directly loaded HBU fuel previously approved by the NRC, specifically, the time limitations associated with the licensing basis ductile-brittle transient temperature limits, as found in SAR section 1.1.1.
The applicant subsequently revised the discussions concerning the HBU fuel contents in the current SAR chapters 2 (Structural) and 3 (Thermal).
3.1 Staff SAR Review The staff conducted the review using the general guidance provided in NUREG-2216 section 3, "Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material.
The staff confirmed that the thermal performance of the Model No. NAC-STC, containing directly loaded HBU fuel, was adequately evaluated for the tests specified under both NCT and HAC of transport.
5 The staff reviewed the applicants proposed changes to SAR chapter 3 and determined that the changes made to the SAR thermal section were primarily editorial in nature. The staff also confirmed that the applicant performed no new thermal performance analyses for the NAC-STC spent fuel transportation system. Therefore, the temperatures reported by the applicant in the previous version of their SAR bounded those in the latest SAR version.
The staff concluded that none of the proposed changes in the SAR would impact the ability of the NAC STC design to meet the thermal performance requirements in 10 CFR Part 71. Based on this conclusion, the staff finds that the NAC-STC package design continues to meet the thermal requirements of 10 CFR Part 71.
3.2 Staff CoC Condition Review The applicant proposed a change to Condition 9(d) of the existing CoC. The existing CoC Condition 9(d) specified that, when fabricating multiple STC packages at a specific fabrication facility using similar fabrication methods, only the first package fabricated needed to undergo the thermal test described in Section 8.1.6 of the NAC-STC SAR. The proposed revised condition removed the language about testing the first STC package fabricated at a specific fabrication facility and specified the general condition for thermal testing as follows: To confirm the NAC-STC heat dissipation design capability, only one first package must be subjected to the thermal acceptance test described in Section 8.1.6 of the NAC-STC Safety Analysis Report.
The staff noted that this change would allow thermal tests performed on previously fabricated NAC-STC packages to be applied to those packages fabricated in the future. The staff discussed the proposed change with the applicant to clarify the intent of the revised condition as documented in the conversation record dated July 25, 2023 (ADAMS Accession No. ML23207A030). Based on the clarification as noted in the conversation record, the staff finds the change to Condition 9(d) of the CoC acceptable.
3.3 Staff Findings Based on a review of the statements and representations in the application, the NRC staff concludes that the thermal design has been adequately described and evaluated, and that the thermal performance of the package meets the thermal requirements of 10 CFR Part 71.
4.0 CONTAINMENT EVALUATION The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the package containment design. Therefore, the staff finds that a new evaluation is not needed.
5.0 SHIELDING EVALUATION The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the package shielding design. Therefore, the staff finds that a new evaluation is not needed.
6
6.0 CRITICALITY EVALUATION
The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the package criticality design. Therefore, the staff finds that a new evaluation is not needed.
7.0 MATERIALS EVALUATION The staff reviewed and evaluated the applicants proposed changes that would credit the flexural rigidity of the fuel pellet as discussed in NUREG-2224 section 2.3.4, Calculation of Cladding Strain Using Factored Cladding-Only Properties. The staff reviewed the changes proposed for SAR sections 2.13.6.15.2, Fuel Rod Assessment for HBU Fuel for 30-foot Side Drop, and 2.13.6.15.4, Fatigue Evaluation for HBU Fuel for Normal Conditions of Transport.
7.1 Vibration The SAR section 2.13.6.15.4 stated that a FE model representing a single WE 17 x 17 fuel rod was used to determine the stress and strain in the fuel cladding during NCT. The applicant performed response spectrum analyses using the transport cask platform response spectra from seven test cases documented in the Data Analysis of ENSA/DOE Rail Cask Tests (SAND2018-13258R, Data Analysis of ENSA/DOE Rail Cask Tests, Spent Fuel and Waste Disposition, U.S. Department of Energy, Spent Fuel and Waste Science and Technology, November 19, 2018)
The applicant reduced the cladding thickness in the FE model by [ ] microns to account for the oxide layer. NUREG-2224 figure 2-5, which the applicant used as guidance, identified oxide layers of 40 to 70 microns for Zircaloy-4. In addition, the applicant cited ORNL/SPR-2020/1744 revision 1, Sister Rod Destructive Examinations (FY21) - Appendix B: Segmentation, Defueling, Metallographic Data and Total Cladding Hydrogen, (Oak Ridge National Laboratory, March 31, 2022) which has the most current data on M5 fuel rods. ORNL/SPR-2020/1744 revision 1 found the oxide layers for M5 fuel rods to measure 8-15 microns on the water side and 7-18 microns on the pellet side. The [ ] micron reduction in cladding thickness chosen by the applicant exceeded the oxide layers identified for Zircaloy-4 in NUREG-2224 and for M5 cladding in ORNL/SPR-2020/1744 revision 1. Therefore, the staff finds the cladding thickness reduction used by the applicant acceptable because it is a conservative assumption that bounds available data on both Zircaloy-4 and M5 fuel rods.
The applicant also applied a flexural rigidity factor of [ ] to the fuel cladding moment of inertia for the cladding only model for both Zircaloy 4 and M5 fuel cladding. The applicant followed the guidance in NUREG-2224 section 2.3.4 in choosing a flexural rigidity factor for the cladding evaluation. The staff requested additional information on the factor chosen for the M5 fuel cladding noting that new data exists that may be informative of the fuel pellet contribution to the M5 fuel flexural response (ORNL/SPR-2020/1780 revision 1, Sister Rod Destructive Examinations (FY21) - Appendix F: Cyclic Integrated Reversible-Bending Fatigue Tests, (Oak Ridge National Laboratory, March 31, 2022). Using actual M5 test data in ORNL/SPR-2020/1780 revision 1 and the flexural rigidity factor based on the NUREG-2224 guidance, the applicant calculated a flexural rigidity ratio of [ ] which bounds the ratio of 1.40 found for Zircaloy 4 in NUREG-2224. Therefore, the staff finds the use of a factor of [ ] for M5 to be acceptable.
7 Based on the above evaluation, the staff finds that the NAC-STC package configuration containing HBU fuel is structurally adequate to meet the requirements of 10 CFR 71.71(c)(5) for the vibration condition.
7.2 Hypothetical Accident Conditions In the PWR High Burnup Fuel Rod 30-ft Side Drop and Fatigue Evaluation for NAC-STC calculation package, the applicant provided the PWR 17 x 17 structural evaluation for the 30-foot side drop accident. The applicant applied a factor of [ ] to the moment of inertia of the cladding-only model for crediting the flexural rigidity of the fuel pellet per NUREG-2224 section 2.3.4, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel. Based on the evaluation above, the staff finds that the NAC-STC containing HBU fuel configuration is structurally adequate in meeting the requirements of 10 CFR 71.73(c)(1) for the free drop tests.
7.3 Fuel Rod Side Drop As described in the SAR section 2.13.6.15.2, the applicant used an ANSYS FE approach to evaluate the fuel rod subject to a bounding side-drop inertia force of [ ]. This value bounded the 51.7g calculated for the NAC-STC with redwood impact limiters and the 50.7g calculated for the balsa impact limiters. Following the guidance in NUREG-2224 section 2.3.4, the applicant took credit for the fuel pellet flexural rigidity by applying a flexural rigidity factor of [ ] to the moment of inertia of the cladding only model. The FEA model calculated a maximum stress of
[ ] ksi as compared to an allowable yield strength of [ ] leading to a factor of safety greater than one. Based on the evaluation above, the staff finds that the HBU fuel assemblies have adequate structural performance such that there will be no permanent deformation of the fuel assemblies due to the HAC tests.
8.0 OPERATING AND MAINTENANCE EVALUATION The changes proposed by the applicant to the STC did not impact the previous materials evaluations that have been performed. The existing analyses in the previous SER for CoC No. 9235 remain applicable for this submittal.
CONDITIONS The CoC includes the following condition(s) of approval:
Condition No. 3(a) was revised to identify the new city name in the address.
Condition No. 5(a)(3)(i) was revised to update the drawing revision numbers.
Condition No. 5(b)(1)(i)(2) was revised to remove text identifying the maximum number of Zirc-4 fuel assemblies per shipment.
Condition No. 5(b)(1)(i)(4) was revised to remove text identifying the maximum number of Zirc-4 fuel assemblies per shipment.
Condition No. 5(b)(2)(i)(2) was revised to remove text identifying the maximum number of Zirc-4 fuel assemblies per shipment.
8 Condition No. 5(b)(2)(i)(4) was revised to remove text identifying the maximum number of Zirc-4 fuel assemblies per shipment.
Condition No. 9(d) was revised to minimize unnecessary performances of thermal tests.
Condition No. 12 was deleted and subsequent conditions were renumbered accordingly.
New Condition No. 15 was added to allow the use of Revision 23 of this certificate until May 31, 2024.
The references section has been updated to include this request.
Minor editorial corrections were made.
CONCLUSIONS Based on the statements and representations contained in the application, as supplemented, and the conditions listed above, the staff concludes that the design have been adequately described and evaluated, and the Model No. STC package meets the requirements of 10 CFR Part 71.
Issued with CoC No. 9235, revision 24, on July 27, 2023.