ML24263A139
| ML24263A139 | |
| Person / Time | |
|---|---|
| Site: | 07103043 |
| Issue date: | 10/09/2024 |
| From: | Tilda Liu Storage and Transportation Licensing Branch |
| To: | Boyle R US Dept of Transportation, Radioactive Materials Branch |
| References | |
| EPID L-2024-DOT-0000 | |
| Download: ML24263A139 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Richard W. Boyle, Chief Radioactive Materials Branch U.S. Department of Transportation 1200 New Jersey Avenue SE Washington, D.C. 20590
SUBJECT:
REQUEST FOR REVALIDATION OF JAPANESE CERTIFICATE OF APPROVAL NO. J/105/AF, DATED SEPTEMBER 13, 2023 - REQUEST FOR SUPPLEMENTAL INFORMATION
Dear Richard Boyle:
By letter dated April 2, 2024 (Agencywide Documents Access and Management System
[ADAMS] Accession No. ML24248A221), the U.S. Department of Transportation (DOT) requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the Japanese Certificate of Approval No. J/105/AF, dated September 13, 2023, for the Model No.
MFC-1 transport package and make a recommendation concerning the revalidation of the package for import and export use.
This letter is to advise you that based on our acceptance review, the application does not contain sufficient technical information. The information needed to begin our review is described in the enclosed request for supplemental information. In order to start our technical review, this information should be provided within 2 weeks from the date of this letter. Upon receiving your responses to this request for supplemental information, the NRC staff will evaluate the information to determine whether the supplementary information is responsive to the NRC staffs concerns.
In addition, your letter provides a non-proprietary version of Enclosure 2, Summary of Changes in Revision 0 of the Safety Analysis Report for the Model MFC-1 Package, but does not provide non-proprietary versions of Enclosures 4, Safety Analysis Report for Model MFC-1 Package, Revision dated December 2023," and Enclosure 5, MNF Fundamental Policy of Quality Management. Please provide non-proprietary versions of Enclosures 4 and 5.
In accordance with Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, requests for withholding, a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records component of the NRCs ADAMS. ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html. The PDR is open by appointment.
To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.
October 9, 2024
R. Boyle Please reference Docket No. 71-3043 and Enterprise Project Identifier No. L-2024-DOT-0000 in future correspondence related to this request. The staff is available to meet to discuss your proposed responses. If you have any questions, please contact me at 404-997-4730.
Sincerely, Tilda Liu, Senior Project Manager Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3043 EPID No. L-2024-DOT-0000
Enclosure:
Request for Supplemental Information Signed by Liu, Tilda on 10/09/24
ML24263A139 OFFICE:
NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME:
BWhite SFigueroa BPatel TBoyce YDiaz-Sanabria TLiu DATE:
9/20/24 9/20/24 9/27/24 9/27/24 10/9/24 10/9/24
Enclosure Request for Supplemental Information Docket No. 71-3043 Model No. MFC-1 Japanese Certificate of Approval J/105/AF By letter dated April 2, 2024 (Agencywide Documents Access and Management System Accession No. ML24248A221), the U.S. Department of Transportation (DOT) requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the Japanese Certificate of Approval No. J/105/AF, dated September 13, 2023, for the Model No. MFC-1 transport package and make a recommendation concerning the revalidation of the package for import and export use.
This request for supplemental information identifies information needed by the NRC staff (the staff) in connection with its review of the application. The staff used the International Atomic Energy Agency (IAEA) Specific Safety Requirements No. 6 (SSR 6), Regulations for the Safe Transport of Radioactive Material," Revision 1 (2018 Edition), in its review of the application.
Structural Evaluation RSI St-1 Provide design fatigue diagrams with an explanation of the associated terminologies and units used in the fatigue evaluation of lifting device components in the safety analysis report (SAR) section II-A.10.9.
The fatigue evaluation of the lifting device components in the SAR table II-A.43 provides the allowable number of cycles, Na, which are calculated from the design fatigue diagram shown in the reference [1], the Japan Society of Mechanical Engineers Standards for Nuclear Power Generation Equipment: Design and Construction Standards (JSME S NC1-2005), including the supplemental edition in 2007. The application did not include either the design fatigue diagrams in the SAR or reference [1]. The staff needs to review these diagrams to verify the structural integrity of the package components under the repetitive design loads.
This information is requested to determine compliance with the requirements in paragraph 613A of the IAEA SSR-6, Rev.1.
Structural Observations Obs-St-1:
Provide a complete evaluation of fatigue for the reusable package structural components for the 60-year period of use that considers the combined effects of all applicable types of accumulated stress cycles in components during normal service conditions.
The staff needs a complete fatigue evaluation that considers the combined effects of all applicable types of stress cycles during normal service, including consideration of the lifting cycles, pressurization cycles, thermal stress cycles and vibration cycles. If certain types of stress cycles are not applicable or negligible for certain components, explain why these are not applicable or are negligible.
The SAR section II-A.4.7 of the application provide an evaluation that demonstrates that package resonance is a not a concern considering package vibration caused by
2 the vehicle during routine condition of transport over the road. However, the staff noted that the fatigue evaluation does not address the potential for fatigue of package components due to accumulation of the vibration cycles effects resulting from the 360 allowed transports of the package during its service-life. Also, the staff could not locate any evaluation of the potential for fatigue of components that accounts for accumulated thermal stress cycles in components. Thermal stress cycles may occur in components due cyclical fluctuation of spatial temperature gradients within components and cyclical temperature fluctuations for assembled or joined components made of dissimilar materials that are dimensionally constrained.
The following provide additional descriptions about some of the package components that may need to consider the combined effects from accumulated stress cycles as provided in the application:
External cylinder: The staff recognizes that external cylinder has been evaluated for lifting cycles in sections (II)-A.10.9 of the SAR. The SAR section II-F, subsection F.1, Aging factors to be considered, states that fatigue evaluation due to internal pressure variations is not performed because the package does not have a containment boundary. However, the staff noted that external cylinder and container dome plate which perform other safety functions, are evaluated for the maximum internal pressure in the SAR section II-A.5.1.3. Therefore, the fatigue evaluation for these components due to internal pressure variation should also be considered. To perform an adequate analytical evaluation that demonstrates sufficient safety margin against fatigue failure of these components, the combined effects of accumulated lifting cycles along with other applicable stress cycles (including consideration of pressurization, thermal and vibration cycle types) on the potential for fatigue in the external cylinder should be considered.
Tie-bolts (Bolts connecting upper and lower halves of external cylinder): The staff recognize that lifting and pressurization cyclic stresses are already considered in the fatigue evaluation of tie-bolts in the SAR section II-A.10.9. However, the staff noted that the effects of thermal and vibration cycles are not considered for the tie-bolts. To perform an adequate analytical evaluation that demonstrates sufficient safety margin against fatigue failure of the tie-bolts, the combined effects of accumulated lifting and pressurization cycles along with other applicable stress cycles (including consideration of thermal and vibration cycle types) on the potential for fatigue in the bolts should be considered.
This information is requested to determine compliance with the requirements in paragraph 613A of the IAEA SSR-6, 2018 Edition.