ML25223A326
| ML25223A326 | |
| Person / Time | |
|---|---|
| Site: | 07103004 |
| Issue date: | 08/11/2025 |
| From: | Garcia-Santos N Division of Fuel Management |
| To: | US Dept of Transportation (DOT), Office of Hazardous Materials Safety |
| References | |
| CAC 001942, CAC 001794, EPID L-2023-DOT-0005, EPID L-2023-DOT-0006 | |
| Download: ML25223A326 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3004 Model No. JMS-87Y-18.5T Package Japanese Competent Authority Certificate (CAC) No. J/2044/B(U)F Table of Contents Page
SUMMARY
....................................................................................................................................1 REGULATORY REQUIREMENTS................................................................................................2 REVALIDATION RECOMMENDATION........................................................................................2
1.0 DESCRIPTION
OF PROPOSED CHANGES....................................................................2 2.0 STRUCTURAL EVALUATION...........................................................................................2 3.0 MATERIALS EVALUATION..............................................................................................7 4.0 THERMAL EVALUATION................................................................................................12 5.0 CONTAINMENT EVALUATION......................................................................................13 6.0 CRITICALITY SAFETY EVALUATION............................................................................13 7.0 SHIELDING EVALUATION.............................................................................................14 8.0 PACKAGE OPERATIONS...............................................................................................14 9.0 QUALITY ASSURANCE PROGRAM..............................................................................15
10.0 REFERENCES
................................................................................................................16 CONCLUSION............................................................................................................................16
SUMMARY
By letter dated April 3, 2023 (Agencywide Documents Access and Management System
[ADAMS] Package Accession Number ML23100A181), and as supplemented on April 13, 2023 (ML23114A256), May 1, 2023 (ML23109A009), June 27, 2023 (ML24043A216), August 28, 2023 (ML23255A102), September 11, 2023 (ML24022A297), September 20, 2023 (ML24046246), February 13, 2024 (ML24046A243), February 14, 2024 (ML24046A242), March 20, 2024 (ML2481A002), and November 15, 2044 (ML24323A035), you requested the U.S.
Nuclear Regulatory Commission (NRC) staff (the staff thereafter) to review the application for the Model No. JMS-87Y-18.5T package and provide a recommendation to revalidate the Japans Competent Authority Certificate (CAC) No. J/2044/B(U)F. The staff reviewed the information provided by the applicant and recommends the revalidation of the package for five years. This document includes the staffs evaluation of DOTs request.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2
OFFICIAL USE ONLY - PROPRIETARY INFORMATION REGULATORY REQUIREMENTS The NRC reviewed the information provided to the DOT by Edlow International Inc. (Edlow or the applicant thereafter) in its application for the Model No. JMS-87Y-18.5T package and its supplements against the regulatory requirements of the International Atomic Energy Agency (IAEA) Safety Standard Series No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, 2018 Edition, Revision 1.
REVALIDATION RECOMMENDATION The staff recommends the revalidation of the Competent Authority Certificate No. J/2044/B(U)F.
The staff proposes adding a condition to the certificate to be issued by U.S. DOT to limit its term to 5 years.
1.0 DESCRIPTION
OF PROPOSED CHANGES The applicant adopted the 2018 Edition of IAEA SSR-6. Also, in the current amendment to the Model No. JMS-87Y-18.5T package, the applicant modified the following fuel element types 1)
JMTR Fuel Followers (HEU) 2)
JMTR Standard Fuel Elements (MEU) 3)
JMTR Fuel Followers (MEU) 4)
JMTR Standard Fuel Elements (LEU) 5)
JMTR Fuel Followers (LEU) 6)
JRR-3 Aluminide Standard Type Fuel Element 7)
JRR-3 Aluminide Follower Type Fuel Element To the following three types:
- 8) JMTR Standard Fuel Elements (MEU)
- 9) JMTR Standard Fuel Elements (LEU)
- 10) JMTR Fuel Followers (LEU)
In addition, the applicant updated the Safety Analysis Report (DOT, 2023) for the package to add aging evaluations to address the requirements of IAEA SSR-6, 2018 Edition, considering a 40-year service life. The following sections include the staffs evaluation of the changes proposed by the applicant.
2.0 STRUCTURAL EVALUATION The staff previously reviewed the structural performance of the Model No. JMS-87Y-18.5T package and recommended revalidation of Japanese Certificate J/150/B(U)F96 (ML040910015) that conformed to the IAEA regulations in TS-R-1, 1996 Edition (IAEA, 1996). A description of the staffs approach to reviewing the changes in the IAEA documents can be found in section 2.2 below.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.1 Description of the Amendment Affecting the Structural Design For this structural review, since the first part of the amendment (i.e., the removal of several fuel elements) did not place any additional demands on the structural components, the structural performance remain unaffected. The staff review is focused on the findings of aging effects on the affected package components and conformance to applicable IAEA SSR-6 (IAEA, 2018a) regulations. This section of the safety evaluation report (SER) documents the staffs reviews, evaluations, and conclusions with respect to the structural integrity of the amended transport package.
2.2 Structural Evaluation of the Revision The staff previously found that the package design had adequate structural integrity to meet the performance standards in IAEA TS-R-1, 1996 Edition. Subsequently, the staff reviewed changes to the IAEA regulations in 2000, 2003, 2005, 2009, and 2012 editions that consisted of clarifications or changes to definitions, updates to general provisions (section III) and restructuring of classification for material and packages (section IV), incorporation of requirements for excepted packages containing uranium hexafluoride (UF6), enhancements to transport modal regulations (e.g., definition, specification), and reconstruction of provisions for the fissile excepted material or packages. These types of changes are either not applicable or have no impact on the structural requirements of the packages that would necessitate reconciliation with the previously revalidated design of the package.
Subsequently, the current 2018 Edition of the IAEA regulations published as the SSR-6, Revision 1, incorporates some notable changes such as the following:
a.
Addition of emergency response to objectives of the Regulations b.
Change in terminology (dose rate instead of radiation level, marking versus mark, etc.)
c.
Introduction of concept of shipment after storage d.
Introduction of SCO-III requirements (shipment of large objects), including definition e.
Deletion of leaching test requirement for LSA-III material f.
Consideration of aging mechanisms during package design g.
Inclusion of a plug-in assessment of individual isolation packages for those containing UF6 Of the above noted changes to the 2018 Edition of the SSR-6 regulations, the addition of requirements for considering aging mechanisms for the package design impacts the structural integrity of the affected package components and remains the focus of the structural evaluation for this package.
2.2.1 Contents Excluded from the package Section (I)-D and Table (I)-A.1 of the SAR and Section 1.0 of this SER describe the contents to be removed as authorized payload in the Model No. JMS-87Y-18.5T package. The staff found, with reasonable assurance, that these changes do not impact the components previously reviewed.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.2.2 Aging Mechanism - Fatigue Paragraph 613A of the IAEA SSR-6 (IAEA, 2018a), requires that aging mechanisms be considered in the design of the package. IAEA Specific Safety Guide No. 26 (SSG-26) (IAEA, 2018b) provides guidance on how to comply with Paragraph 613A in IAEA SSR-6. The Model No. JMS-87Y-18.5T is a Type B(U)F (fissile material) package and intended for repeated shipment use. Therefore, in accordance with the guidance provided in Paragraphs 613A.1 and 613A.3 of the IAEA SSG-26, the package needs to be evaluated for the effects of aging mechanisms during the design phase to demonstrate compliance with the transport regulations.
The application provides the aging evaluation in section (II)-F of the SAR to address the requirements of Paragraph 613A of IAEA SSR-6 in the package design and operations. The following documents the staffs reviews and evaluations with respect to the structural integrity of the package components affected by fatigue, one of the aging mechanisms. Section 3.0, Materials Evaluation, of this SER provides an evaluation due to aging mechanisms of the materials of construction used to fabricate the package.
The reusable package components including the lifting device for the package body and the lid, the containment device, tie-down attachments, and the skid lifting device attachments are subject to repetitive loads due to handling of the package and components, differential pressure, temperature changes, and vibration loads during transport as applicable through the expected service life of the package. As a result, the application evaluates fatigue for the lifting device attachments, containment device components, and tie-down attachments in the SAR
[(DOT, 2023), (DOT, 2023)], and supplemented responses [(EDL, 2023a), (EDL, 2023b) and (DOT, 2024)] to the NRCs requests for additional information (RAIs).
2.2.2.1 Handling (Lifting) Cycle For the fatigue evaluation of the stainless-steel lifting device attachments to the package body and the lid, the applicant conservatively considered 4,000 loading cycles (estimated 800 cycles at 20 lifts per year) over the 40-year service life, as stated in section (II)-A.4.4.3.3 and (II) Table F.2 of the SAR and supplemented responses to the RAI. The applicant showed that the loading cycles is lower than the [Information withheld per 10 CFR 2.390] allowable number of cycles associated with the maximum cyclic stress [Information withheld per 10 CFR 2.390] in the body lug based on the established design fatigue curve [Information withheld per 10 CFR 2.390] of the SAR for the austenitic type of stainless-steel material. [Information withheld per 10 CFR 2.390] Therefore, the staff finds this evaluation conservative and acceptable since the allowable number of loading cycles associated with the maximum cyclic stress is greater than the estimated loading cycles over the specified service life.
For the fatigue evaluation of the carbon steel skid lifting device, the applicant conservatively considered 4,000 loading cycles (estimated 800 cycles at 20 lifts per year) over the 40-year service life, as stated in section (II)-A.10.6.4(3) of the SAR. The applicant showed that the maximum cyclic stress [Information withheld per 10 CFR 2.390] in the skid lifting lug is lower than the fatigue strength [Information withheld per 10 CFR 2.390] associated with the 4,000 loading cycles based on the established design fatigue curve [Information withheld per 10 CFR 2.390] for the carbon steel material. The staff notes that the evaluation is based on a conservative estimate of loading cycles, and determines this evaluation to be acceptable since the allowable number of loading cycles associated with the maximum cyclic stress is greater than the estimated loading cycles over the specified service life.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.2.2.2 Thermal and Pressurization Cycle For the fatigue evaluation of the containment device (i.e., package body, lid, and the connecting bolts) due to repetitive loads (i.e., pressure, thermal) under normal conditions of transport (NCT), the applicant estimated 40 loading cycles based on the 40-year life and one transport per year. However, the applicant considered 1,000 loading cycles over the 40-year life in the fatigue evaluation presented in the SAR section (II).A.10.5 and Table F.2, and supplemented responses to the RAI. For the containment device components, the applicant showed that the margin of safety [Information withheld per 10 CFR 2.390], where FS is defined as a ratio of the allowable to estimated number of stress cycles] in each component for the governing NCT case is positive, based on the established design fatigue curve [Information withheld per 10 CFR 2.390]of the SAR for the austenitic type of stainless-steel and high strength bolts material respectively. In particular, for the most critical component, the lid-bolts, the applicant showed that the estimated [Information withheld per 10 CFR 2.390] loading cycles is lower than the allowable number of loading cycles [Information withheld per 10 CFR 2.390] associated with the repetitive peak stress intensity based on the established design fatigue curve in
[Information withheld per 10 CFR 2.390] the SAR for the high strength bolt material.
The staff reviewed the applicants evaluation and the relevant SAR sections and finds that although the estimated number of pressurization and thermal stress cycles for the containment device considered is appropriate for the condition when fuel content is loaded for the transportation, the number of thermal stress cycles due to daily changes in spatial temperature exceed 1,000 cycles considered during the 40-year service life. The staff also finds that sufficient conservatism exists in the applicants fatigue evaluation, because approximately 70% of the stresses in the lid-bolts are due to installation torque that are associated with a smaller number of torquing cycles over the service life than that considered in the fatigue evaluation. Thus, the staff determines the applicants fatigue assessment for the containment device and the lid-bolts due to pressure and temperature stress cycles to be acceptable.
2.2.2.3 Vibration Cycle The applicant presented a strength analysis of the tie-down attachments in the SAR, section (II).A.4.5, and a fatigue assessment for the vibration cycles in the supplemented responses to the RAI. For the fatigue evaluation of the tie-down attachment (tie-down lug) to the package body of the package, the applicant showed that the maximum cyclic stresses in the tie-down lug is lower than the allowable fatigue strength corresponding to the [Information withheld per 10 CFR 2.390] loading cycles based on the established design fatigue curve in [Information withheld per 10 CFR 2.390] the SAR for the austenitic stainless-steel material. The maximum cyclic stress in the tie-down lug considered in this evaluation is based on the application of static accelerations (10G forward/backward, 5G left/right, 2G up/down) forces on the package during transport.
The staff reviewed the applicants evaluation and notes that the cyclic stress for the fatigue evaluation is based on conservative strength analysis acceleration values (10G forward/backward, 5G left/right, 2G up/down), as compared to the acceleration values for the fatigue analysis imparted by rail transport mode ( +/- 0.3G forward/backward, +/- 0.4G left/right,
+/- 0.3g) per Table IV.3 of IAEA SSG-26 (IAEA, 2018b), and by truck transport mode (1.5G forward/backward, 1.17G left/right, 2G up/down) per Table III of NUREG/CR-0128, Shock and Vibrations Environment for Large Shipping Container during Truck Transport. The staff also notes that there will not be any load amplification due to vibration during transport as demonstrated by the applicant in the SAR section (II)-A.4.7, since there is no potential for
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 6
OFFICIAL USE ONLY - PROPRIETARY INFORMATION vibration resonance during transport due to the relatively high natural frequency of the package.
The staff finds this evaluation acceptable based on the conservative assumptions made by the applicant, and the cyclic stress amplitude being lower than the stainless-steel material endurance limit [Information withheld per 10 CFR 2.390].
2.2.2.4 Combined Effects of Fatigue Cycles The staff notes that a fatigue evaluation combining the effects of all applicable types of accumulated stress cycles (i.e., lifting, pressurization, thermal, and vibration) in the affected components during normal service conditions needs to show that fatigue failure will not occur.
The combined effects of applicable cycle types in the relevant package components are typically evaluated using a traditional method of superposition and calculating a cumulative usage factor (i.e., summation of ratios of the estimated over the allowable number of cycles for each applicable cycle type) for an affected component. If the cumulative usage factor for a critical package component combining applicable stress cycle types is less than 1.0, then it demonstrates that a fatigue failure of an evaluated component will not occur over the specified service life.
The staff reviewed the applicants evaluation of package components for different types of stress cycles. Based on this review, the staff finds that the most critical components are the package body and the lid-bolts. Although the lifting lug locations [Information withheld per 10 CFR 2.390] at the top of the package body and the tie-down lug locations [Information withheld per 10 CFR 2.390] on the exterior side near the top of the package body are not at the same radial locations, they are conservatively considered to have an overlapping effect on the package body from lifting and vibration type stress cycles. Conservatively combining fatigue usage factors for lifting, vibration, temperature and pressure stress cycles, the staff estimates the cumulative usage factor in the package body to be less than 1.0. For the lid-bolts, bolt installation torque cycle, pressure and temperature cycles, and lid self-mass acceleration due to vibration during transport need to be considered in evaluating combined effects. Based on the results of the individual stress cycle type, the staff conservatively estimates the cumulative usage factor for the tie-bolts to be less than 1.0. With this assessment and due to inherent conservatism in the stress analyses for individual stress cycle type, the staff finds the fatigue evaluation to be acceptable for combined effects on the applicable package components.
Based on the above assessments, the staff concludes that there is no adverse impact on the structural adequacy of the reusable important-to-safety package components due to the aging effects from various applicable fatigue cycle types.
2.2.3 Structural Evaluation Under the Routine Condition of Transport, Normal Condition of Transport (NCT) and Accident Condition of Transport (ACT)
The applicants previously reviewed evaluations for the other routine condition of transport (RCT), NCT, and ACT are bounding or remain unaffected by the changes described above for this amendment.
2.3 Evaluation Findings
Based on a review of the statements and representations contained in the application, the staff finds, with reasonable assurance, that the design of the Model No. JMS-87Y-18.5T package provided adequate structural capacity to meet the requirements of IAEA SSR-6, 2018 Edition, Revision 1.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 7
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.0 MATERIALS EVALUATION 3.1 Regulatory Requirements The purpose of the materials evaluation is to verify that the performance of the materials used to build the package components meets the requirements of IAEA SSR-6, 2018 Edition, Paragraph No. 613A to ensure that the package design considered and appropriately evaluated aging mechanisms. A summary of the staffs materials evaluation is provided below.
3.2 Evaluation The staff reviewed the adequacy of the package materials of the Model No. JMS-87Y-18.5T against the IAEA SSR-6, 2018 Edition, requirements related to the materials performance of the package, and the ability of the package design to meet such requirements. The applicant did not request changes to the packaging design. The applicant requested approval to remove JRR-3 Aluminide fuel (Standard and Follower), JMTR Standard and Follower (HEU) and JMTR Fuel Followers (MEU), for revalidation of the transportation package for import and export use. The package is designed to meet the general use, function, and testing requirements specified in IAEA SSR-6, 2018 Edition. The package is made of welded stainless steel and consists of a canister body holding a fuel basket containing fuel elements.
3.2.1 Deletion of Registered Contents The staff reviewed the deletion of the registered contents above and finds the change acceptable as it does not change the materials safety analyses of the previously approved fuels.
3.2.2 Adoption of 2018 Edition of IAEA SSR-6 and Considerations of Aging The staff previously reviewed the materials performance of the JMS-87Y-18.5T package and its conformance to IAEA TS-R-1, 1996 Edition, in the recommended revalidation of Japanese Certificate No. J/150/B(U)F96 (ML040910015). The staff reviewed the changes since the 1996 edition through the current 2018 edition and found that the only change applicable for the material performance review was the consideration of aging mechanisms for the package design.
The applicant added Section F, Consideration of Aging of Nuclear Fuel Package to Chapter II:
Safety Analysis of the Packages, to address aging of the transport container.
In Section F.1, the applicant described the conditions of use and various requirements that must be satisfied for storage, before transport, during transport, and after transport. While in storage, periodic inspection is performed at least once a year based on Chapter III, Maintenance of transport containers and handling methods of nuclear fuel packages, of the SAR. Before transport, a pre-shipment inspection based on Chapter III is conducted. During transport, the package is securely tied to withstand shock and vibration expected during the 2-month period of transport. After transport, a visual inspection is conducted in a controlled area to confirm the integrity of the transport container. The SAR did not include the aging of O-rings as they are replaced with each transport. In addition, the aging of package contents is not considered because these are loaded at each shipment.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION In Section F.2, the applicant provided evaluations of stainless steel (body, lid, and basket,
[Information withheld per 10 CFR 2.390]-Aluminum alloy), Aluminum alloy (spacer), and Wood [Information withheld per 10 CFR 2.390], and the aging factors considered were heat, radiation, chemical changes, and fatigue. The applicant concluded that under the conditions of use expected during the planned period, compliance with the technical criteria would not be affected by aging.
The applicant described the requirements that the inspection program would comply with:
[Information withheld per 10 CFR 2.390]
The applicant stated that internal guidelines for visual inspection include specific targets and acceptance criteria for the components:
[Information withheld per 10 CFR 2.390]
Personnel performing the inspections will be qualified to the Japan Society of Mechanical Engineers [Information withheld per 10 CFR 2.390], are certified under Japan Materials Testing Reactor [Information withheld per 10 CFR 2.390], and undergo annual eye examinations [Information withheld per 10 CFR 2.390].
The applicant stated that visual inspections are performed in accordance with JSME S NC1-2005 [Information withheld per 10 CFR 2.390]. The inspector will also ensure that adequate lighting is provided [Information withheld per 10 CFR 2.390].
The applicant described surface cleaning requirements to ensure that bare metal visual inspections of component surfaces are capable of detecting surface flaws.
The applicant described flaw evaluation and corrective actions in their internal guidelines including specific criteria for replacement of vent plugs and drain valves, O-rings, gaskets, and bolts and nuts. The applicant also provided an example of a 2022 corrective action case
[Information withheld per 10 CFR 2.390].
The staff reviewed the applicants package maintenance activities including visual inspections, screening and evaluation of visual indications, and corrective actions such as component repairs and replacements, and found that:
a.
Aging mechanisms and effects identified by the applicant are applicable to the package component materials and operating environments.
b.
Examination techniques are sufficient to detect aging mechanisms and effects for the package components.
c.
Maintenance program includes actions to address and mitigate aging mechanisms and effects as required.
d.
The applicant has provided personnel qualification requirements for staff that conduct inspections.
The staff found that the maintenance activities are adequate to manage the effects of aging in metallic package components that would see long-term use, such that the package components are capable of performing their requisite safety functions throughout the period of use.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.2.3 Abrasion Per the applicant, wear can occur in parts that experience intermittent relative motion or frequent manipulation. The applicant states that the package does not have dynamic structures or bearings, or other structures that could deteriorate due to wear. Parts that may experience wear include the bolts and bend valves that tighten the sealing boundaries of the transport package (lid and drain/vent valves). The applicant identified nine areas [Information withheld per 10 CFR 2.390].
The applicant detailed the inspection procedures, frequency and criteria for replacement in Chapter III, Operation and Maintenance Methods of Packages, B.1.1, Visual Inspection, B.2, Leakage Test, and B.4, Maintenance of Valves Gaskets, etc. of Packaging.
The staff reviewed the evaluation for abrasion performed by the applicant and found that the applicants assessment is limited to abrasion as a result of assembly and disassembly operations, but the visual inspections discussed in the previous section are sufficient to identify and characterize abrasion on any other surface of the package. Therefore, the staff finds that the applicant adequately demonstrated that the inspection and maintenance procedures will manage any aging effects that could affect functionality under routine conditions of transport.
3.2.4 Stainless Steel Heat - The applicant stated that the thermal analysis indicated that the maximum temperature would peak at [Information withheld per 10 CFR 2.390] during transportation. Since this is below 425°C, the temperature at which deformation due to creep would occur1, the applicant does not expect any aging effects due to heat during the period of use. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to heat that could affect functionality under routine conditions of transport.
Radiation - The applicant stated that the maximum neutron irradiation dose during the period of use is [Information withheld per 10 CFR 2.390], which is less than the dose of 1016 n/cm2 that may cause embrittlement2. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to radiation that could affect functionality under routine conditions of transport.
Chemical Changes - The applicant stated that the depth of corrosion that could occur in the air is estimated to be 0.001mm per year with a maximum of 0.04mm during the period of use. The applicant stated that this is a negligible amount of corrosion compared to the thicknesses of the component materials [Information withheld per 10 CFR 2.390] for the package body. The staff reviewed the evaluation for corrosion performed by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to corrosion that could affect functionality under routine conditions of transport.
1 Transportation Technology Advisory Board, "Measures to Ensure Safety of Post-Storage Transportation for Interim Storage of Spent Fuel" (2010).
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 10 OFFICIAL USE ONLY - PROPRIETARY INFORMATION Fatigue - The detailed fatigue evaluation is discussed in the structural evaluation Section 2.2.2 of this SER.
3.2.5
[Information withheld per 10 CFR 2.390]-Aluminum Alloy)
Heat - The applicant stated that the results of the thermal analysis indicated that the maximum temperature during transportation is [Information withheld per 10 CFR 2.390], which is well below the melting temperature of 2,450°C, therefore the applicant does not expect any aging effects due to heat during the period of use. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to heat that could affect functionality under routine conditions of transport.
Radiation - The applicant stated that the neutron irradiation dose during the period of use is [Information withheld per 10 CFR 2.390], conservatively assuming a period of use of 100 years. Since the loss of 10B is estimated to be 0.0043%, this means the loss of 10B due to neutron irradiation is negligible. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to radiation that could affect functionality under routine conditions of transport.
Chemical Changes - The applicant stated that corrosion does not occur because it is in a sealed space within the basket dividers (stainless steel) and does not contact the outside air. The staff reviewed the evaluation for corrosion performed by the applicant and finds that the applicant adequately demonstrated that this material will not undergo any aging effects due to corrosion during the period of use that could affect functionality under routine conditions of transport.
3.2.6 Aluminum Alloy (spacer)
Heat - The applicant stated that the results of the thermal analysis indicated that the maximum temperature during transportation is [Information withheld per 10 CFR 2.390], which is below the melting temperature of 660°C, therefore the applicant does not expect aging effects due to heat during the period of use. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to heat that could affect functionality under routine conditions of transport.
Radiation - The applicant stated that the maximum neutron irradiation dose during the period of use is [Information withheld per 10 CFR 2.390], which is less than the dose of 1021 n/cm2 that may cause embrittlement2. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to radiation that could affect functionality under routine conditions of transport.
Chemical Changes - The applicant stated that as aluminum alloys form an oxide film on its surface, they are not susceptible to corrosion during the period of use. In addition, the material is checked for abnormalities in its appearance before shipment. The staff notes that environments encountered during loading and unloading operations, including spent fuel pool water with boric acid, will not result in degradation of the protective oxide layer or a significant increase in the aluminum alloy corrosion rate. The staff reviewed
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 11 OFFICIAL USE ONLY - PROPRIETARY INFORMATION the evaluation for corrosion performed by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to corrosion that could affect functionality under routine conditions of transport.
3.2.7 Wood (Fir-Plywood)
Heat - The applicant stated that the calculated maximum temperature of the surface of the package is [Information withheld per 10 CFR 2.390] during transportation based on the thermal analysis. The maximum temperature inside the shock absorber (wood) is
[Information withheld per 10 CFR 2.390] during transportation, based on the thermal analysis. The applicant stated that there is a lack of literature that can be directly referenced regarding the strength degradation of wood due to heat in a high temperature environment.
The applicant stated that the average temperature data of the shock absorber of another package with a track record of transportation of spent fuel showed the temperature to range between 40°C and 70°C.
The temperature of the surface of the package during actual transportation has been measured at about 40°C. Based on the analysis performed on the surface of the package and the temperature during actual transportation, the temperature inside the shock absorber during actual transportation is estimated to be below 40°C. Crush strength and density measurements were also performed on the wood specimens from the package and concluded that the energy absorption performance of the wood was not affected at this temperature, nor a deterioration in performance.
The applicant also provided a study published by Shikoku Electric Power Co., Inc, on the thermal deterioration of shock absorbing material (wood). The staff reviewed the study and found that based on the conditions under which packages are normally used, thermal deterioration of the shock absorbing material did not affect package performance.
The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to heat that could affect functionality under routine conditions of transport.
Radiation - The applicant stated that the neutron dose expected during the period of use is conservatively estimated to be about [Information withheld per 10 CFR 2.390],
and the gamma radiation dose is about [Information withheld per 10 CFR 2.390],
which is lower than the radiation dose of 3 MGy (neutron) and 0.1 MGy (gamma) that can cause microstructural changes like embrittlement2. The staff performed an independent review to verify the neutron and gamma dose limits for embrittlement and finds the dose to be below the amount that could cause embrittlement. The staff reviewed the information provided by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to radiation that could affect functionality under routine conditions of transport.
2 Gilbert Gedeon, P.E., Wood as An Engineering Materials: Mechanical Properties of Wood Course No:S04-019
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 12 OFFICIAL USE ONLY - PROPRIETARY INFORMATION Chemical Changes - The applicant stated that corrosion does not occur because it is in a sealed space within the shock absorber liner (stainless steel), moisture was controlled in the wood prior to sealing, and the wood does not contact the outside air. The staff reviewed the evaluation for corrosion performed by the applicant and finds that the applicant adequately demonstrated that this material will not undergo aging effects due to corrosion that could affect functionality under routine conditions of transport.
The staff reviewed Section F and the evaluation of aging factors including heat, radiation, chemical changes, and fatigue and confirmed that the package meets the requirements of IAEA SSR-6, 2018 Edition, Paragraph 613A.
3.3 Evaluation Findings
The staff finds, with reasonable assurance, that the mechanical and heat transfer properties of the materials used in the fabrication of the Model No. JMS-87Y-18.5T package area are acceptable to transport the requested three types of JMTR fuel. The staff bases this finding on the JMTR fuels being previously approved and that the proposed changes do not expose the packaging materials to temperatures, radiation dose, or other environmental exposure that were not previously evaluated by the staff. The staff concludes that the applicant adequately described and evaluated the materials used in the Model No. JMS-87Y-18.5T package and that the package meets the requirements of IAEA SSR-6, 2018 Edition, Paragraph 613A.
4.0 THERMAL EVALUATION 4.1 Regulatory Requirements The purpose of the thermal evaluation is to verify that the package design meets the requirements of IAEA SSR-6, 2018 Edition, for this revalidation. The requested changes do not have any impacts on the current thermal performance of the package. A summary of the staffs review is provided below.
4.1 Normal Conditions of Transport The removal of certain contents from the package Certificate, as requested by the applicant, will have no effect on the thermal performance of the package and continues to be bounded under normal conditions of transport; therefore, the package meets the requirements of IAEA SSR-6, 2018 Edition.
4.2 Accident Conditions of Transport (ACT)
The thermal performance of the JMS-87Y-18.5T package with the requested contents removed, will have no effect on the thermal performance of the package and continues to be bounded under ACT; therefore, the package meets the requirements of IAEA SSR-6, 2018 Edition.
4.3 Evaluation Findings
Based on review of the statements and representations in the application, the staff finds with reasonable assurance that the thermal evaluation for the Model No. JMS-87Y-18.5T meets the requirements of the IAEA SSR-6, 2018 Edition, for normal and ACT.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 13 OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.0 CONTAINMENT EVALUATION 5.1 Regulatory Requirements The purpose of the containment evaluation is to verify that the Model No. JMS-87Y-18.5T package design meets the requirements of IAEA SSR-6, 2018 Edition, when evaluated for normal and ACT. The requested changes do not have any impacts on the current thermal performance of the package. A summary of the staffs containment evaluation is provided below.
5.2 Evaluation The removal of certain contents from the package Certificate, as requested by the applicant, will have no effect on the thermal performance of the package and continues to be bounded under normal and ACT; therefore, the package meets the requirements of IAEA SSR-6, 2018 Edition.
5.3 Evaluation Findings
Based on a review of the statements and representations contained in the application, the staff finds with reasonable assurance that the containment design and evaluation for the Model No.
JMS-87Y-18.5T meets the requirements of the IAEA SSR-6, 2018 Edition.
6.0 CRITICALITY SAFETY EVALUATION 6.1 Regulatory Requirements The purpose of the criticality safety evaluation is to verify that the package design meets the requirements of IAEA SSR-6, 2018 Edition. A summary of the staffs criticality safety evaluation is provided below.
6.2 Evaluation The applicant requested a U.S. revalidation of the CAC for the Model No. JMS-87Y-18.5T package to the requirements of IAEA SSR-6, 2018 Edition. The NRC previously revalidated the certificate for this package in 2004 (ML040910015). This applicant removed previously approved contents from the CAC and added aging management procedures as required by IAEA SSR-6, 2018 Edition.
The applicants process for performing the criticality analysis did not change when removing contents from the CAC. The applicant modeled each fuel type and assumed all gaps within the package were filled with water and minimum thickness for the neutron absorber. The water density with maximum reactivity was found to determine optimum moderation. This analysis resulted in a maximum system keff + 3 of 0.864, including all biases and uncertainties, which is below the 0.95 acceptance criteria. The removal of the other previously registered contents does not change the conclusions of this analysis.
The aging management protocol that affects criticality safety is the depletion of the neutron absorber throughout the lifetime of the package. For the analysis of the depletion of the neutron absorber by neutron irradiation, the applicant extended the time period analyzed to 100 years
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 14 OFFICIAL USE ONLY - PROPRIETARY INFORMATION and determined that the depletion of the neutron absorber [Information withheld per 10 CFR 2.390] will have minimal effect on criticality safety throughout the lifetime of the package.
The staff reviewed the certificate for the Model No. JMS-87Y-18.5T package, as well as the applicants initial assumptions, model configurations, analyses, and results in the SAR. The staff finds that the applicant has identified the most reactive configuration of the Model No. JMS-87Y-18.5T package with the requested contents, and that the criticality results are conservative and demonstrate the package and package arrays will be subcritical.
6.3 Evaluation Findings
Based on a review of the statements and representations contained in the application, the staff finds with reasonable assurance that the criticality safety evaluation for the model No. JMS-87Y-18.5T meets the requirements of the IAEA SSR-6, 2018 Edition.
7.0 SHIELDING EVALUATION 7.1 Regulatory Requirements The purpose of the shielding evaluation is to verify that the package design meets the requirements of IAEA SSR-6, 2018 Edition. A summary of the staffs shielding evaluation is provided below.
7.2 Evaluation The staff previously revalidated the certificate for this package in 2004 (ML040910015). The change within the application is the deletion of different fuel element types and their shielding results. Although the applicant removed fuel element types as allowable contents from the certificate, the applicant did not change any of the other analyses of the three other fuel types.
The fuel type with the greatest combined gamma ray and neutron dose equivalent rates is the 30 JMTR standard fuel element with a dose equivalent rate one meter away from the package surface of 0.083 mSv/h and on the package surface of 0.998 mSv/h. These calculated dose rates satisfy the SSR-6 requirements of 0.1 mSv/h or less and 2 mSv/h or less, respectively.
7.3 Evaluation Findings
Based on a review of the statements and representations contained in the application, the staff finds with reasonable assurance that the shielding evaluation for the Model No. JMS-87Y-18.5T meets the requirements of the IAEA SSR-6, 2018 Edition.
8.0 PACKAGE OPERATIONS Package operations are discussed and evaluated in different sections of this SER.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 15 OFFICIAL USE ONLY - PROPRIETARY INFORMATION 9.0 QUALITY ASSURANCE PROGRAM 9.1 Regulatory Requirements The purpose of the quality assurance (QA) review is to verify that the package design meets the requirements of the IAEA SSR-6, 2018 Edition. The staff reviewed the description of the QA program for the Model No. JMS-87Y-18.5T package against the standards in the IAEA SSR-6.
9.2 Evaluation The applicant developed and described a QA program for activities associated with transportation packaging components important to safety. Those activities include design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use. The applicants description of the QA program (i.e., management system and compliance assurance programs in IAEA SSR-6, 2018 Edition) meets the requirements of the applicable IAEA SSR-6, 2018 Edition. The staff finds the QA program description acceptable, since it allows implementation of the associated QA program for the design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use of the Model No. JMS-87Y-18.5T transportation package.
The staff finds, with reasonable assurance, that the QA program for the JMS-87Y-18.5T transportation packaging meets the requirements in IAEA SSR-6, 2018 Edition by encompassing the following:
1) design controls, 2) materials and services procurement controls, 3) records and document controls, 4) fabrication controls, 5) nonconformance and corrective actions controls, 6) an audit program, and 7) operations or programs controls, as appropriate.
These controls are adequate to ensure that the package will allow safe transport of the radioactive material authorized in this approval.
9.3 Evaluation Findings
Based on review of the statements and representations in the Model No. JMS-87Y-18.5T package application and as discussed in this SER section, the staff has reasonable assurance that the package meets the requirements in IAEA SSR-6, 2018 Edition. The staff recommends revalidation of Japanese CAC No. J/2044/B(U)F.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 16 OFFICIAL USE ONLY - PROPRIETARY INFORMATION
10.0 REFERENCES
(IAEA, 1996)
International Atomic Energy Agency No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, 1996 Edition.
(IAEA, 2018a)
International Atomic Energy Agency No. SSR-6, Regulations for the Safe Transport of Radioactive Material, 2018 Edition.
(IAEA, 2018b)
International Atomic Energy Agency No. SSG-26, Revision 1, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (2018 Edition).
(DOT, 2023)
Richard W. Boyle, U.S. Department of Transportation (DOT), letter to Director, Division of Fuel Management, U.S. Nuclear Regulatory Commission (NRC), April 3, 2023, ML23100A181.
(EDL, 2023a)
Russell Neely, Edlow International Company, Letter to Rick Boyle, U.S.
Department of Transportation (DOT), September 11, 2023, ML24022A297.
(EDL, 2023b)
Russell Neely, Edlow International Company, Letter to Rick Boyle, U.S.
Department of Transportation (DOT), September 14, 2023, ML24023A031.
(DOT, 2024)
Rick Boyle, U.S. Department of Transportation (DOT), Letter to Ms.
Norma Garcia Santos U.S. Nuclear Regulatory Commission (NRC),
November 15, 2024, ML24323A035.
CONCLUSION Based on the statements and representations in the information provided by DOT and the applicant, the staff recommends the revalidation of the Japanese CAC No. J/2044/B(U)F, Model No. JMS-87Y-18.5T package.