ML20189A268
| ML20189A268 | |
| Person / Time | |
|---|---|
| Site: | 07105797 |
| Issue date: | 07/07/2020 |
| From: | John Mckirgan Storage and Transportation Licensing Branch |
| To: | Shuler J US Dept of Energy, Office of Packaging and Transportation |
| Saverot P | |
| Shared Package | |
| ML20189A267 | List: |
| References | |
| EPID L-2020-LLA-0119 | |
| Download: ML20189A268 (6) | |
Text
T. Tate UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 7, 2020 Dr. James M. Shuler Manager, DOE Packaging Certification Program U.S. Department of Energy Office of Packaging and Transportation EM-4.24, 270CC - Rm 3113 Washington, DC 20585
SUBJECT:
CERTIFICATE OF COMPLIANCE NO. 5797, REVISION NO. 21, FOR THE MODEL NO. INNER HFIR UNIRRADIATED FUEL ELEMENT SHIPPING CONTAINER, AND OUTER HFIR UNIRRADIATED FUEL ELEMENT SHIPPING CONTAINER
Dear Dr. Shuler:
As requested by your application dated May 28, 2020, enclosed is Certificate of Compliance No.
5797, Revision No. 21, for the Model No. Inner HFIR unirradiated fuel element shipping container and outer HFIR unirradiated fuel element shipping container package. Changes made to the enclosed certificate are indicated by vertical lines in the margin. The U.S. Nuclear Regulatory Commission staffs safety evaluation report is also enclosed.
The approval constitutes authority to use the package for shipment of radioactive material and for the package to be shipped in accordance with the provisions of Title 49 of the Code of Federal Regulations 173.471.
If you have any questions regarding this certificate, please contact Pierre Saverot of my staff at 301-415-7505.
Sincerely, John McKirgan, Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-5797 EPID - L-2020-LLA-0119 Enclosures 1: Safety Evaluation Report 2: Certificate of Compliance cc w/encl:
R. Boyle, Department of Transportation John B. McKirgan Digitally signed by John B.
McKirgan Date: 2020.07.07 14:29:32 -04'00'
- via email OFFICE NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME:
- PSaverot
- JChang
- YDiaz-Sanabria
- SFigueroa
- JMcKirgan DATE:
06/25/2020 06/25/2020 06/25/2020 07/06/2020 07/7/2020
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Safety Evaluation Report U.S. Department of Energy Docket No. 71-5797 Model No. Inner HFIR Unirradiated Fuel Element Shipping Container, and Outer HFIR Unirradiated Fuel Element Shipping Container BACKGROUND The High Flux Isotope Reactor (HFIR) Unirradiated Fuel Element Shipping Container is a type B(U)F package, with contents described as U3O8-Al cermet, enriched up to 95% in U235, and clad in aluminum. U234 and U236 have both a corresponding A2 value of 6.0 x10-3 TBq, with the remaining U235 and U238 being unlimited with respect to A2, as shown in 10 CFR Part 71 Appendix A, Table A-1.
During a recent review by ORNL of the Safety Analysis Report (SAR) of the Model No. Inner HFIR Unirradiated Fuel Element Shipping Container, and Outer HFIR Unirradiated Fuel Element Shipping Container (Docket No. 71-5797), non-conservative values were identified in association with the hypothetical accident conditions (HAC) thermal test damage conditions for the side plates of both the inner and outer HFIR fuel elements. ORNL identified that the pre-machined plate thickness was used, instead of the final thickness, in the analysis for the side plates of both the inner and outer HFIR fuel elements. As such, the reduced thickness increases both the exposure time to melt the fuel cladding and the fraction of exposed/unclad fuel, and consequently the A2 results.
The Department of Energy (DOE) notified the NRC by letter dated March 4, 2020, of this calculation error in the thermal analysis. Based on the staffs review of the applicants evaluation, the NRC issued a Letter of Authorization, dated March 18, 2020, which authorized the use of the HFIR package until July 31, 2020.
By letter dated May 28, 2020, DOE requested an amendment to the Certificate of Compliance (CoC) and submitted a corrected HAC thermal analysis.
EVALUATION As stated in Enclosure 1 of the amendment request, Evaluation of the Error and Proposed Path Forward, the applicant incorrectly assumed the pre-machined side plate thickness instead of the final thickness for the HAC thermal analysis to predict the percentage of the fuel-bearing material that would be melted if a fuel assembly was directly exposed to the HAC fire without the benefit of the container. The applicant used the pre-machined side plate thicknesses of 0.377 inch for the inner fuel assembly and 0.408 inch for the outer fuel assembly and predicted that approximately 13% of the fuel-bearing material for both inner fuel assembly and outer fuel assembly would be melted under the HAC fire.
DOE/ORNL subsequently determined that this calculation error is not safety significant and that the results from the revised HAC thermal analysis do not change any of the CoC conditions because no loss of radioactive contents exceeds one A2 per week under HAC.
Material Properties and Component Specifications The applicant stated in Appendix B of the SAR Rev. 12 that the average specific heat for the side plates, made of 6061 aluminum, is about 0.247 Btu/lbm-°F from 169°F to 1080°F, and the fusion heat of the aluminum cladding is 171 Btu/lbm. The applicant referred to a report by M.
W. Wendel and D. G. Morris, High Flux Isotope Reactor System RELAPS Input Model, Report ORNL/TM-11647, Draft 12-20-90 to state that: (a) the content for the inner side plates is 77 wt% aluminum with an average specific heat of 0.218 Btu/lbm-°F from 169°F to 1220°F and an average density of 176 lbm/ft3, and (b) the content for the outer side plates is 71 wt% aluminum with an average specific heat of 0.207 Btu/lbm-°F from 169° F to 1220°F and an average density of 186 lbm/ft3.
The staff reviewed this ORNL report, and the aluminum material source book, and confirmed that the thermal properties of the 6061 aluminum, used by the applicant in the revised thermal analysis, are acceptable.
Normal Conditions of Transport (NCT)
The applicant stated in Section 3.4 of the SAR Rev. 12 that (a) the previous assessments and over 20 years of in-service operations demonstrate that, for the thinner side plates of the inner and outer fuel assemblies, the package is adequate for normal conditions of transport (NCT) and (b) there are no thermal gradients sufficient to cause any significant thermal stress in the package.
The staff reviewed the SAR Rev. 12 and agrees that the use of the thinner side plates with final/correct thicknesses for the inner and outer fuel assemblies has no significant effects either on the package thermal performance or on the maximum thermal stresses under NCT. This is because the thinner side plates will not cause significant changes in the temperatures of the package metallic components under NCT and, therefore, the previous NCT thermal evaluation and its stress calculations remain acceptable and in compliance with 10 CFR 71.51(a)(1).
Hypothetical Accident Conditions (HAC)
Melt of the Fuel Side Plates for Inner Fuel Assembly and Outer Fuel Assembly The applicant described the revised HAC thermal analysis and the related parameters in Appendix B of the SAR Rev. 12:
the thinner side plates with final/correct thicknesses (0.105 inch for the inner fuel assembly and 0.128 inch for the outer fuel assembly), and the HAC initial temperature of 169°F for the package, the melting point of 1080°F for the fuel cladding material 6061 aluminum (outer cladding) and the melting point of 1220°F for the side plates (77 wt% aluminum) of the inner fuel assembly and for the side plates (71 wt% aluminum) of the outer fuel assembly.
The applicant first derived the nominal melt rate of 1.90 inch/hour into clad material 6061 aluminum, and then the melting rates of 1.63 inch/hour for the inner side plate and 1.64 inch/hour for the outer side plate, and finally the melts of 0.668 inch for the inner side plate and the 0.672 inch for the outer side plate. Correspondingly, the calculated melted fractions of the fuel-bearing material are 21.8% for the inner fuel assembly and 24.4% for the outer fuel assembly, as shown in Appendix B of the SAR Rev. 12.
The staff reviewed the revised HAC thermal analysis for melting of the fuel inner side plate and the fuel outer side plate and finds the assumptions, parameters, and methodology used in the analysis to be reasonable. The staff confirmed that less than 25% of the fuel-bearing material of a HFIR fuel assembly would melt if the fuel assembly were exposed to a 1475°F fire for 30 minutes.
Fractions for the Unclad Fuel Assembly The applicant described the revised HAC thermal analysis and the related parameters in Chapter 4 of the SAR Rev. 12:
the radionuclides of U234, U235, U236, and U238 contained in unirradiated HFIR fuel, the specific activities of 2.3 x 10-4 TBq/gram for U234 and 2.4 x 10-6 TBq/gram for U236 and the unlimited A2 values for U235 and U238, per 10 CFR 71, Appendix A, and the loading limits of 2630 grams and 6880 grams, respectively, for U235 in the inner fuel assembly and the outer fuel assembly.
The applicant stated in Chapter 4 of the SAR Rev. 12 that the HFIR fuel is highly enriched U235 with a weight 93.164% and therefore is characterized as a mixture of isotopes U234, U235, U236 and U238 to determine the A2 value for the mixture. The applicant provided the data of specific activity, weight %, and fraction of activity for these isotopes in Appendix B of the SAR Rev. 12.
The staff referred to 10 CFR Part 71 Appendix A (Table A-1) and agrees that the data (e.g., A2 values, specific activities, and weight percentages of U234, U235, U236 and U238 used in the revised HAC thermal analysis are consistent with U234, U235, U236 and U238 numbers listed in Table A-1 of the Part 71 Appendix A and are acceptable for the revised calculation.
The applicant calculated an A2 of 1.65 x 10-1 Curies (Ci) for the uranium fuel mixture and the uranium fuel activities of 1.83 x 10-1 Ci and 4.79 x 10-1 Ci, respectively, for the inner fuel assembly and the outer fuel assembly. As shown in Appendix B of the SAR Rev. 12, the applicant, using these activities, concludes that an amount of fuel bearing material exceeding 90% of uranium in the inner fuel assembly or exceeding 34% of uranium in the outer fuel assembly constitutes a quantity exceeding A2 value. The applicant concluded, in SAR Amendment Enclosure 2, that the amount of fuel which melts in the regulatory fire is more than the previously assumed (13% increasing to approximately 25%); however, based on revised calculations for the amount of fuel which can melt prior to releasing an A2 value in one week, the applicant found that there is still a positive margin of safety against the release of radioactive material.
The staff reviewed the applicants calculation and agrees that greater than 90% of the fuel-bearing material in the inner fuel element or greater than 34% of the fuel-bearing material in the outer fuel element must melt before the regulatory limit defined in the 10 CFR 71.51(a)(2) would be exceeded. Therefore, the staff determines that the HAC fire would not result in a release of an A2 value in one week, in compliance with 10 CFR 71.51(a)(2).
EVALUATION FINDINGS The fraction of unclad fuel for the thinner side plate (after-machining, with a final thickness of 0.105 inch) is 21.8% for the inner fuel assembly and is below the HAC limit of 90% which corresponds to the limit defined in 10 CFR 71.51(a)(2).
The fraction of unclad fuel for the thinner side plate (after-machining, with a final thickness of 0.128 inch) is 24.4% for the outer fuel assembly and is below the HAC limit of 34% which corresponds to the limit defined in 10 CFR 71.51(a)(2).
Based on these findings listed above, the staff has reasonable assurance that, for both inner fuel assembly and outer fuel assembly, there will be no escape of radioactive material exceeding a total amount of A2 in one week under the HAC fire and, thus, the package is in compliance with 10 CFR 71.51(a)(2). The CoC No. 71-5797 does not need to be modified and stays as issued.
CONDITIONS The following changes were made to the CoC:
Item No. 3(b) identifies now correctly the application, as supplemented.
Condition No. 11 extends the use of Revision No. 20 of the CoC by approximately one more year.
The References section of the certificate was updated to include the supplement dated May 2020 (Revision No. 12 of the application -pages changes).
CONCLUSION Based on the statements and representations in the application, the staff finds that these changes do not affect the ability of the Model No. Inner HFIR Unirradiated Fuel Element Shipping Container, and Outer HFIR Unirradiated Fuel Element Shipping Container package to meet the requirements of 10 CFR Part 71.
Issued with CoC No. 5797, Revision No. 21.