ML25223A273
| ML25223A273 | |
| Person / Time | |
|---|---|
| Site: | 07103035 |
| Issue date: | 08/11/2025 |
| From: | Garcia-Santos N, Yoira Diaz-Sanabria Division of Fuel Management |
| To: | Cheng K, Vierling R US Dept of Transportation (DOT), Office of Hazardous Materials Safety |
| References | |
| 01942, 01794, EPID L-202023-DOT-0005 | |
| Download: ML25223A273 (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. 71-3035 Model No. JRC-80Y-20T Package Japanese Competent Authority Certificate (CAC) No. J/2045/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................................................................................................11 5.0 CONTAINMENT EVALUATION......................................................................................12 6.0 CRITICALITY SAFETY EVALUATION............................................................................12 7.0 SHIELDING EVALUATION.............................................................................................13 8.0 PACKAGE OPERATIONS...............................................................................................14 9.0 QUALITY ASSURANCE PROGRAM..............................................................................14
10.0 REFERENCES
................................................................................................................15 CONCLUSION............................................................................................................................15
SUMMARY
By letter dated April 3, 2023 (Agencywide Documents Access and Management System
[ADAMS] Package Accession Number ML23115A059), and as supplemented on April 13, 2023 (ML23115A075), 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, 2024 (ML24323A035), you requested the U.S. Nuclear Regulatory Commission (NRC) staff (the staff thereafter) to review the application for the Model No. JRC-80Y-20T package and provide a recommendation to revalidate the Japanese Competent Authority Certificate (CAC) No. J/2045/B(U)F (ML23115A083). The staff recommends the revalidation of the CAC No. No. J/2045/B(U)F for a period of five years. This document includes the staffs evaluation of DOTs request.
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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. JRC-80Y-20T 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 CAC No. JRC-80Y-20T. 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 revision to the Model No. JRC-80Y-20T package, the applicant modified the following fuel element types:
1)
JRR-3 Silicide Standard Type Fuel Element 2)
JRR-3 Silicide Follower Type Fuel Element 3)
JRR-3 Aluminide Standard Type Fuel Element 4)
JRR-3 Aluminide Follower Type Fuel Element 5)
JRR-4 Low Enriched Fuel Element 6)
JRR-4 High Enriched Instrumentation Fuel Element To the following three types:
7)
JRR-3 Silicide Standard Type Fuel Element 8)
JRR-3 Silicide Follower Type Fuel Element 9)
JRR-3 MNU Fuel In addition, the applicant updated the Safety Analysis Report (DOT, 2023a) for the package to add aging evaluations to address the requirements of IAEA SSR-6, 2018 Edition, considering a 70-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 JRC-80Y-20T package and recommended revalidation of Japanese Certificates J/61/B(U)F-96, Revision 3 (ML19213A037) that conformed to the IAEA regulations in SSR-6, 2012 Edition. A description of the staff s approach to reviewing the changes in the IAEA documents can be found in section 2.2 below.
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 staffs review is focused on the findings of aging effects on
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 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 Amendment The staff previously found that the package design had adequate structural integrity to meet the performance standards in SSR-6, 2012 Edition. Subsequently, the staff reviewed changes to the IAEA regulations in the 2018 Edition, which 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. JRC-80Y-20T package. The staff found, with reasonable assurance, that these changes do not impact the components previously reviewed.
2.2.2 Aging Mechanism - Fatigue Paragraph 613A of 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), Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, (IAEA, 2018b) provides guidance on how to comply with Paragraph 613A in the IAEA SSR-6. The Model No.
JRC80Y-20T is a Type B(U)F, fissile material, package intended for repeated shipment.
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 in the IAEA SSR-6.
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 staff reviewed and evaluated the structural integrity of the package components affected by fatigue, one of the aging mechanisms. Section 3.0, Materials Evaluation, of this SER provides a detailed safety evaluation of other aging mechanisms.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION The reusable package components including the lifting device for the packages body and the lid, and the containment device are subject to repetitive loads due to handling of the package/components, differential pressure, temperature changes and vibration loads during transport as applicable through the expected service life of the package. As a result, the applicant evaluated fatigue of the lifting device attachments, tie-down attachments and containment device components in the SAR [(DOT, 2023a)], and supplemented responses
[(DOT, 2023b), (EDL, 2023) 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 attachment (i.e., lifting lug) for the package body and the lid, the applicant conservatively considered 10,000 loading cycles (estimated 7,000 cycles at 100 lifts per year) over the 70-year service life, as stated in sections (II)-A.4.4.2.1.3 and (II)-A.4.4.2.2.4 of the SAR. The applicant showed that the number of stress cycles considered for this evaluation is lower than the allowable number of stress cycles
[Information withheld per 10 CFR 2.390] corresponding to the calculated maximum cyclic stress for the body lifting lug from the established design fatigue curve [Information withheld per 10 CFR 2.390]. The staff finds that the lug cyclic stress considered for fatigue evaluation is based on a conservative lifted weight (i.e. three times the weight of the package). Also, the estimated loading cycles considered during the service life is conservative. Because of these conservatisms and since the allowable number of loading cycles is shown to be greater than the estimated loading cycles during the service life, the staff determines this fatigue evaluation to be acceptable.
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 during transport, the applicant evaluated the most critical component, the lid bolts, for the thermal [Information withheld per 10 CFR 2.390] pressure loads considering 300 operating cycles (estimated 280 cycles at 4 times handling of sealing device per year) over the 70-year service life, as stated in section (II)- A.5.1.3 and Table F.2 of the SAR. The applicant showed that the estimated number of stress cycles during the service life is lower than the allowable number of stress cycles [Information withheld per 10 CFR 2.390] corresponding to the calculated maximum cyclic stress for the bolt per the established design fatigue curve [Information withheld per 10 CFR 2.390], for high strength steel bolting per the ASME BPVC,Section III, Division 1, Appendices, 1986.
During the review of the applicants evaluation, the staff noted 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 the 300 cycles considered during the 70-year service life. However, even considering 2.6 x 104 number of thermal stress cycles (i.e., 365 cycles per year for 70 years) due to daily temperature variation, the allowable number of cycles corresponding to the bolt thermal cyclic stress exceeds the estimated number of thermal stress cycles during the service life. Therefore, the staff finds the applicants fatigue assessment for the containment device and the lid-bolts, due to pressure and temperature stress cycles, to be acceptable.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.2.2.3 Vibration Cycle The package is tied down by tightening of the fin bolts compressing the tie-down block on the shipping skid against the package body bottom fins shoes as shown in the SAR (II).Fig. A.27.
The horizontal tie-down forces are distributed through the bottom seven fins each at the front and rear in the direction of travel, the bottom four fins each at the right and left in the direction perpendicular to the travel, and through the same total 22 bottom fins in the vertical direction.
The applicant provided a strength analysis of the tie-down attachments to the bottom fins in the SAR [Information withheld per 10 CFR 2.390], and a fatigue assessment for vibration cycles in the supplemented responses to the RAI. For the fatigue evaluation of the tie-down attachments (bottom fins) to the package body of the package, the applicant demonstrated that the maximum cyclic stress in the bottom fins is lower than the allowable fatigue strength corresponding to the 1.0 X 106 loading cycles based on the established design fatigue curve
[Information withheld per 10 CFR 2.390] of the Model No. JRC-80Y-20T SAR for the austenitic stainless-steel material. The maximum cyclic stress in the tie-down bottom fins considered in this evaluation is based on applying 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 vibration resonance during transport due to the relatively high natural frequency of the package.
The staff considers 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 of approximately [Information withheld per 10 CFR 2.390] of the Model JRC-80Y-20T SAR.
The applicant has presented a strength analysis of the tie-down attachments in the SAR, section (II).A.4.5, and a fatigue assessment for 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]of 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
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 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 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 of approximately [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 need 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 [Information withheld per 10 CFR 2.390], 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 for package components for different types of stress cycle. 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 (total 2) on the exterior wall of the package body at the top and the tie-down attachment locations [Information withheld per 10 CFR 2.390] on the bottom fins of the package are far apart such that the effects of the pressure and temperature cycles with either the effects of lifting cycles or vibration cycles need to be combined, both do not need to be combined. The lid-bolts are the most critical components. Therefore, the lid-bolts installation torque cycle, pressure and temperature cycles, and lid self-mass acceleration due to vibration cycles during transport need to be considered in evaluating the combined effects on the lid-bolts. Based on the results of the individual stress cycle type, the staff conservatively estimated the cumulative usage factor for the tie-bolts to
[Information withheld per 10 CFR 2.390] [Information withheld per 10 CFR 2.390]. 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 all 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. JRC-80Y-20T package
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION provides adequate structural capacity to meet the requirements of IAEA SSR-6, 2018 Edition, Revision 1.
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. JRC-80Y-20T 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 package is designed to meet the general use, function, and testing requirements specified in SSR-6, 2018 Edition. The package is made of welded stainless steel and consists of a package body holding a fuel basket containing fuel elements.
3.2.1 Deletion of Registered Contents The staff reviewed the deletion of registered contents as described in section 1.0 of this SER and finds the changes 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 JRC-80Y-20T package and its conformance to IAEA SSR-6, 2012 Edition, in the recommended revalidation of Japanese Certificate No. J/61/B(U)F96 (ML19213A037). The staff reviewed the changes since the 2012 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 to be 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 transportation. In addition, 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), Boral (boron carbide-aluminum alloy), and aluminum alloy (spacer), 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 (JSME) S NC1-2005, Standards for Nuclear Power Plant Equipment Design and Construction Standards, are certified under Japan Materials Testing Reactor Information withheld per 10 CFR 2.390]1, and undergo annual eye examinations in accordance with the Industrial Safety and Health Act (Japan).
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 by use of a hard-hat headlamp plus flashlight.
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 where abnormalities were discovered and addressed including the initial determination, the repair, reporting to the Nuclear Regulation Authority to confirm that the appropriate measures were taken, and the final inspection after repair.
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.
1 Internal certification that defines standards, internal procedures, specifications, equipment, and operational and maintenance requirements of the package being inspected
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 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.
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 four areas including:
[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, Visual Inspection, B.3, Leakage Inspection, and B.10, Maintenance of Valves Gaskets, etc. of containment vessel.
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 fuel elements would peak at [Information withheld per 10 CFR 2.390] during transportation and the fuel basket center axis would peak at 200°C. Since this is below 425°C, the temperature at which deformation due to creep would occur2, 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 0.07mm 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 2 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 CFR 2.390] for the transport container 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.
Fatigue - The 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 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 0.00013%, 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 us that could affect functionality under routine conditions of transport.
3.2.6 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 660°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 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 any aging effects due to radiation that could affect functionality under routine conditions of transport.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 11 OFFICIAL USE ONLY - PROPRIETARY INFORMATION 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 inspected for abnormalities 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 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.
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, requirement 613A.
3.3 Evaluation Findings
The staff finds that the mechanical and heat transfer properties of the materials used in the fabrication of the Model JRC-80Y-20T package area are acceptable to transport the requested content. The staff bases this finding on the JRR-3 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 JRC-80Y-20T package and that the package meets the requirements of IAEA SSR-6, 2018 Edition.
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-62018 Edition.
4.2 Accident Conditions of Transport (ACT)
The thermal performance of the JRC-80Y-20T 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-62018 Edition.
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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. JRC-80Y-20T meets the requirements of the IAEA SSR-6, 2018 Edition, for normal and ACT.
5.0 CONTAINMENT EVALUATION 5.1 Regulatory Requirements The purpose of the containment evaluation is to verify that the Model No. JRC-80Y-20T 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 packages 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.
JRC-80Y-20T 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. JRC-80Y-20T package to the requirements of IAEA SSR-6, 2018 Edition. NRC previously revalidated the CoC for this package in 2019 (ML19213A037). 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 criticality analysis did not change when removing contents from the CoC. The applicant modeled each fuel type and assumed all gaps within the package are filled with water along with only taking [Information withheld per 10 CFR 2.390] credit for the boron neutron absorber. The water density with maximum reactivity was found to determine optimum
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 13 OFFICIAL USE ONLY - PROPRIETARY INFORMATION moderation. This analysis resulted in a maximum system keff + 3 of 0.873, 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 possibility of degradation 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 used the conservative neutron source determined within the shielding section of the SAR. Even after extending the time period analyzed to 100 years, the applicant determined the depletion of the neutron absorber would
[Information withheld per 10 CFR 2.390] have minimal effect on criticality safety throughout the lifetime of the package.
The staff reviewed the certificate for the Model No. JRC-80Y-20T 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. JRC-80Y-20T package with the requested contents, and that the criticality results are conservative and demonstrate the package and package arrays will be subcritical. Therefore, the staff finds with reasonable assurance that the package, with the requested contents, will meet the criticality safety requirements of IAEA SSR-6, 2018 Edition.
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. JRC-80Y-20T 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 Japanese certificate for the Model No. JRC-80Y-20T package was revalidated in in 2019 (ML19213A037). The only changes are described in section 1.0 of this SER.
Although the applicant removed four fuel types as allowable contents from the certificate, the applicant still used the fuel type with the greatest gamma and neutron source to determine bounding external dose rates. The strongest gamma source occurring from the JRR-3 silicide type fuel, and the neutron source from a removed fuel type which has been renamed as Fuel Element B. Fuel Element B has the same characteristics as the fuel type used in the previous revalidation in 2017, just under a generic name. The name change does not affect the analysis of the source term or dose rate and maintains conservative estimates of the external dose rates of the package, which remain below the dose rate limits in SSR-6, 2018 Edition.
The staff reviewed the certificate for the Model No. JRC-80Y-20T package, as well as the applicants initial assumptions, model configurations, analyses, and results in the application.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 14 OFFICIAL USE ONLY - PROPRIETARY INFORMATION The staff finds with reasonable assurance that the package, with the requested contents will meet the radiation dose rate requirements of IAEA SSR-6, 2018 Edition.
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. JRC-80Y-20T 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.
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. JRC-80Y-20T package against the standards in the IAEA SSR-6, 2018 Edition.
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. JRC-80Y-20T transportation package.
The staff finds, with reasonable assurance, that the QA program for the JRC-80Y-20T 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
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 15 OFFICIAL USE ONLY - PROPRIETARY INFORMATION 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. JRC-80Y-20T 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/2045/B(U)F.
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, 2023a)
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, ML23115A075.
(DOT, 2023b)
Richard W. Boyle, U.S. Department of Transportation (DOT), letter to Project Manager, Division of Fuel Management, U.S. Nuclear Regulatory Commission (NRC), June 27, 2023, ML031780469.
(EDL, 2023)
Russell Neely, Edlow International Company, Letter to Rick Boyle, U.S.
Department of Transportation (DOT), September 11, 2023, ML24022A297.
(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/2045/B(U)F, Model No. JRC-80Y-20T package.