ML20197A183

Request for Additional Information for Review of the Certificate of Compliance No. 3083

ML20197A183
Person / Time
Site:	Orano USA
Issue date:	07/24/2020
From:	Garcia-Santos N Storage and Transportation Licensing Branch
To:	Boyle R US Dept of Transportation, Office of Hazardous Materials Safety
NGSantos NMSS/DFM/STL 415.6999
References
EPID L-2020-LLA-0011
Download: ML20197A183 (6)

Text

July 24 ,2020 Mr. Richard W. Boyle, Chief Sciences Branch Division of Engineering and Research Office of Hazardous Materials Safety U.S. Department of Transportation 1200 New Jersey Ave., S.E.

Washington, D.C. 20590

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR REVIEW OF THE CERTIFICATE OF COMPLIANCE NO. 3083 (EPID L-2020-LLA-0011)

Dear Mr. Boyle:

By letter dated December 12, 2019 [Agencywide Documents Access and Management System (ADAMS) Accession No. ML20052D258], the U.S. Department of Transportation requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the French Approval Certificate Number F/347/AF-96, Revision Fs, Model No. FCC-3 transport package, and makes a recommendation concerning the revalidation of the package for import and export use.

In connection with our review, we need the information identified in the enclosure to this letter.

To assist us in scheduling the staff review of your response, we request that you provide this information by September 4, 2020. Please inform us at your earliest convenience, but no later than August 14, 2020, if you are not able to provide the information by that date. If you are unable to provide a response by September 4, 2020, our review may be delayed.

Please reference Docket No. 71-3083 and EPID L-2020-LLA-0011 in future correspondence related to this request. The staff is available to meet to discuss your proposed responses if desired. If you have any questions regarding this matter, I may be contacted at (301) 415-6999 or ngs@nrc.gov.

Sincerely, Digitally signed by Norma Garcia Norma Garcia Santos Santos Date: 2020.07.24 06:57:55 -04'00' Norma García Santos, Project Manager Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3083 EPID L-2020-LLA-0011

Enclosure:

Request for Additional Information

Request for Additional Information U.S. Department of Transportation French Approval Certificate No. F/347/AF-96, Revision Fs Docket No. 71-3083 Certificate of Compliance No. 3083 Model No. FCC-3 By letter dated December 12, 2019 [Agencywide Documents Access and Management System (ADAMS) Accession No. ML20052D258], the U.S. Department of Transportation requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the French Approval Certificate Number F/347/AF-96, Revision Fs, Model No. FCC-3 transport package, and make a recommendation concerning the revalidation of the package for import and export use.

This request for additional information (RAI) identifies information needed by the NRC staff (the staff) in connection with its review of the application. The staff used International Atomic Energy Agency (IAEA) Specific Safety Requirements No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, 2012 Edition, in its review of the application.

The RAI describes information needed by the staff to complete its review of the application and to determine whether the applicant has demonstrated compliance with the regulatory requirements of SSR-6.

CRITICALITY SAFETY (Cr)

RAI-Cr-1 Provide the chemical composition and the concentration of the chromium oxides that will be present within the uranium dioxide (UO2) pellets to be shipped in the Model No. FCC-3.

Certificate of the French Competent Authority No. F/347/AF-96, Revision Fs, for the Model No. FCC-3 package, notes that, for all contents, the UO2 pellets may contain chromium oxides. The mass of uranium is dependent on the chromium oxides. The application does not include specific information about the chemical composition of the chromium oxides and concentrations that will exist within the UO2. The staff needs this information to ensure that the fissile material to be transported will be subcritical.

This information is needed to determine compliance with the requirements in Paragraph 673 of SSR-6.

RAI-Cr-2 Demonstrate and justify that ignoring the inclusion of chromium oxide within the UO2 is conservative (i.e., reduces reactivity) with respect to criticality safety.

The following criticality analysis files state that the chromium oxides are ignored in the criticality analysis:

Enclosure

• 160 PUBLIC Encl 4 FCC 3 Ap 2.5-1 FFDC00817 EN, Safety/Criticality Study of FCC3 and FCC4 Containers 15x15 and 17x17 Assemblies 17x17XL, 17x17 EPR', 16x16 and 18x18 Assemblies (095 SUNSI Encl 3 FCC 3 Ap 2.5-1 FFDC00817 EN) that represents Content Nos. 1 and 2, for the FCC-3, and

• 162 PUBLIC Encl 4 FCC 3 Ap 2.5-2 FFDC01046EN, Safety/Criticality Study of FCC3 and FCC4 Containers Transport of Fuel Rods in Boxes (097 SUNSI Encl 3 FCC 3 Ap 2.5-2 FFDC01046EN) that represents Content Nos. 5, 6, 7, and 8 for the FCC-3.

The aforementioned files state the following:

The addition into the pellets of capturing compounds (chromium oxide for instance) in quantities close to those of the impurities is ignored in the study of criticality as their presence reduces the reactivity of the assemblies.

In addition, the following criticality file does not mention how chromium oxide is represented within the criticality analysis:

164 PUBLIC Encl 4 FCC 3 Ap 2.5-3 FFDC01106EN (099 SUNSI Encl 3 FCC 3 Ap 2.5-3 FFDC01106-A E0.pdf) that represents Contents Nos. 3 and 4 for the FCC-3.

The staff has knowledge about studies (see References 1 and 2 of this question) to support criticality safety analyses for other packages involving UO2 fuel with chromium oxides that show that the reactivity effect of a specific quantity of chromium oxide may be statistically insignificant for a small number of rods.

However, the staff requests justification that the proposed level of chromium oxide reduces reactivity, if all rods contain chromium oxides.

References:

(1) Letter from the NRC to W. Stilwell, Westinghouse Electric Company, Special Authorization for a One Time Shipment of the Model No.

Traveller Package, December 20, 2018, ADAMS Accession No. ML18354B136.

(2) Letter from NRC to T. Tate, Framatome Inc., Revision No. 11 of Certificate of Compliance No. 9319, for the Model Nos. MAP-12 and MAP-13 Transportation Packages, January 10, 2019, ADAMS Accession No. ML19011A013.

This information is needed to determine compliance with the requirements in Paragraph 673 of the IAEA SSR-6, 2012 Edition.

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RAI-Cr-3 Justify the modeling of the assembly with respect to the pitch of the assembly rods (see fissile cross section from Table 2 of the App. 2.5-1) under normal conditions of transport and accident conditions in the criticality evaluation for Content Nos. 1 and 2 of the FCC-3 in Appendix 2.5-1 of the safety analysis report (SAR).

For Content Nos. 1 and 2, the application does not specify the pitch of the rods in the content specifications within the certificate. Although the staff recognizes that 1/3 of the assembly is modeled with maximum pitch under accident conditions, it requests that the applicant justify that the pitch for the entire length of the assembly is appropriate for both normal and accident conditions because the pitch of the assembly affects the criticality calculations.

This information is needed to determine compliance with the requirements in Paragraphs 682, 684, and 685 of SSR-6.

RAI-Cr-4 Justify that the criticality code bias and uncertainty, determined for the CRISTAL V1.1 code package, are applicable or bounding for the uncertainty for the CRISTAL V0.2 code package.

In App. 2.5-6 of the SAR, the applicant provided an evaluation of the bias and uncertainty for codes and cross sections used to perform criticality evaluations.

This document notes that the applicant determined the uncertainty using the codes from the CRISTAL V1.1 code package.

App. 2.5-3 of the SAR contains the criticality evaluations for Content Nos. 3 and 4 of the FCC-3 and notes that the applicant used the CRISTAL V0.2 code package. As discussed in Section 3 of App. 2.5-3 of the SAR, the applicant is applying the same code uncertainty that was determined for the CRISTAL V1.1 code package in Appendix 2.5-6 to the results that used the CRISTAL V0.2 code package.

This information is needed to determine compliance with the requirements in Paragraphs 682, 684, and 685 of the IAEA SSR-6, 2012 Edition.

RAI-Cr-5 The staff requests that the applicant state the cross sections used to perform the criticality evaluations in App. 2.5-1, 2.5-2 and 2.5-3 of the safety analysis report.

If they are not the same as those used within the benchmarking evaluations for the criticality codes in App. 2.5-6, justify the use of these cross sections.

Appendix 2.5-6 of the SAR show that the applicant used the JEF 2.2 cross section library to perform the benchmarking of the CRISTAL V1.1 criticality code package. App. 2.5-1, 2.5-2 and 2.5-3 of the SAR that documents the criticality evaluations for the FCC-3 does not state what cross section library the applicant used in its evaluations.

The applicant needs to verify that it used the same cross section library in its criticality evaluations as it did to perform the benchmarking analyses, or provide justification that the criticality code uncertainty established in App. 2.5-6 is applicable to the cross sections used in the FCC-3 criticality evaluations.

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The staff requests this information so that it can determine compliance with Paragraphs 682, 684, and 685 of the IAEA SSR-6, 2012 Edition.

RAI-Cr-6 Explain how the uncertainty from the MORET code, associated with the Monte Carlo method, is applied to the results of the criticality analyses for the FCC-3.

In Section 3 of App. 2.5-1, 2.5-2 and 2.5-3 of the SAR, the applicant notes that the uncertainty value of the CRISTAL code (as determined by App. 2.5-6 of the SAR) is added to all of the reactivity values in the report. However, the staff requests clarification on how the uncertainty of the MORET code associated with the Monte Carlo method for each calculation is applied to each of the results.

Typically, applicants will add 2 to the nominal keff value calculated. The staff cannot find in the application how this was treated in App. 2.5-1, 2.5-2 and 2.5-3 of the SAR.

This information is needed to determine compliance with the requirements in Paragraphs 682, 684, and 685 of the IAEA SSR-6, 2012 Edition.

RAI-Cr-7 Provide the following information related to assumptions for the array calculations under accident conditions in the FCC-3 criticality evaluations for Content Nos 1 and 2:

a) a demonstration showing that the condition of empty space between packages is more reactive than a condition with hydrogenous material between packages; and b) the assumed spacing between the packages and justification for this spacing.

In Section 5.3.2 of App. 2.5-1 of the SAR, the applicant notes that there is empty space between the packages. Paragraph 685(a) of the SSR-6 notes that criticality analysis for arrays under accident conditions shall have hydrogenous moderation between the packages. The staff requests that the applicant provide information demonstrating compliance consistent with Paragraph 685(a) of SSR-6, or update the criticality analysis to justify the applicants assumption of the most reactive packages for the array condition under accident conditions.

The staff did not find a discussion in App. 2.5-1 of the SAR on the spacing of the packages for the array condition under accident conditions for the criticality analysis for the FCC-3 Content Nos. 1 and 2 in App. 2.5-1 of the SAR.

This information is needed to determine compliance with the requirements in Paragraph 685(a) of SSR-6.

RAI-Cr-8 Provide a justification related to the assumptions used to represent the neutron absorber within the criticality analysis are appropriate to incorporate the effects of the thermal test in SSR-6 Paragraph 728 and that the neutron absorber would not melt under credible accident conditions. Include the following information as part of your response:

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a) Justification for the conditions of the package and assumptions used to determine the effects of thermal tests on the neutron absorber resin; b) Documentation justifying the glass transition temperature of the neutron absorber resin; and c) Explanation of the statement that a boron gradient was formed within the material in Appendix 2.2-4 of the SAR.

The applicant demonstrated the effects of the thermal tests described in SSR-6 Paragraph 728 on the neutron absorber resin in the following appendices to the application:

• Appendix 2.2-4 of the SAR, which discusses the direct flame and furnace fire tests on the neutron absorber resin, and

• Appendix 2.2-1 of the SAR, which includes a thermal analysis using the STAR-CD code.

The condition of the neutron absorber within the criticality evaluation (10% loss of boron and loss of hydrogen in the outer 10mm) is more consistent with the results of the tests in Appendix 2.2-4 rather than Appendix 2.2-1, which shows that the neutron absorber resin could melt. The staff recognizes that the differences in these effects are likely due to the assumptions of the condition of the package used in the tests/modeling.

The staff requests that the applicant provide additional information justifying the conditions that are most appropriate for use in the thermal evaluation so that the staff can determine if the effects on the resin within the criticality evaluation are appropriate. The staff is primarily concerned with the possibility of the neutron absorber resin melting and whether this is more limiting from a criticality perspective.

This information is needed to determine compliance with the requirements in Paragraphs 673, 682, and 685 of SSR-6.

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