ML25227A182

Enclosure - Safety Evaluation Report, Docket No. 71-3043 Revalidation of Model No. MFC-1 Transport Package, Japanese Competent Authority Certificate No. J/105/AF

ML25227A182
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Site:	07103043
Issue date:	09/05/2025
From:	Storage and Transportation Licensing Branch
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EPID L-2024-DOT-0000
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Enclosure SAFETY EVALUATION REPORT DOCKET NO. 71-3043 REVALIDATION OF MODEL NO. MFC-1 TRANSPORT PACKAGE JAPANESE COMPETENT AUTHORITY CERTIFICATE NO. J/105/AF

SUMMARY

By letter dated April 2, 2024 (Agencywide Documents Access and Management System Accession No. ML24248A221), as supplemented by letter dated October 23, 2024, (ML24306A041), and its response dated May 15, 2025 (ML25140A904) to the U.S. Nuclear Regulatory Commission (NRC) staffs request for additional information (RAI), dated May 14, 2025 (ML25113A127), the U.S. Department of Transportation (DOT) requested that the NRC staff perform a review of the Competent Authority of Japan, Certification for Approval of Package Design for Transport of Radioactive Materials, Identification Mark J/105/AF, dated September 13, 2023, for the Model No. MFC-1 transport package. Specifically, DOT requested the NRC to make a recommendation concerning the revalidation of the Model No. MFC-1 transport package for import and export use against the requirements found in International Atomic Energy Agency (IAEA) Safety Standard Series No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, Revision 1, 2018 Edition.

In support of its request, by letter dated April 2, 2024, DOT provided the following documents to the NRC for revalidation review:

TN Americas LLC (TN) Letter to DOT, dated December 3, 2023, Materials Package Design Certificate USA/0535/AF-96, Revalidation

• : English version of Japanese Approval Certificate of a Package Design, Number J/105/AF

• : Summary of changes in the MFC-1 Safety Analysis Report (SAR)

• : Revision comparison table for MFC-1 SAR Rev. August 2016 and Rev.

December 2023 [proprietary version]

• : Proprietary version of the Mitsubishi Nuclear Fuel Co., Ltd. (MNF) Safety Analysis Report for MFC-1, Rev. December 2023 [proprietary version]

• : MNF Fundamental Policy of Quality Management, Rev. December 2023

[proprietary version]

The NRC staff reviewed DOTs request (i.e., the application), as supplemented on October 23, 2024, and May 15, 2025, against the requirements in IAEA SSR-6, Revision 1, 2018 Edition (hereafter generally referred to as SSR-6). Based upon the NRC review of the statements and representations contained in the documents listed above, as supplemented, and for the reasons stated in this safety evaluation report (SER), the NRC recommends revalidation of Japanese Competent Authority Certificate of Approval No. J/105/AF, dated September 13, 2023, for the Model No. MFC-1 package, with the condition that the U.S. revalidation term is limited to 5 years.

2 1.0 GENERAL INFORMATION The Model No. MFC-1 is a Type A fissile package designed for the transport of unirradiated Pressurized Water Reactor (PWR) fuel.

1.1 Packaging A description of the package is provided in the Model MFC-1 SAR. The package consists of a main body, lifting trunnions, lid and bottom plug, shock absorber, and seals. The fuel assembly rods provide containment of the radioactive material transported. Therefore, the MFC-1 package does not utilize a containment system. The contents allowed by the Japanese certificate are PWR fuel assemblies fabricated either from uranium dioxide (UO2) or gadolinia UO2. The allowable uranium-235 (235U) enrichment in the fuel assemblies is 5 weight percent (wt.%) or less.

1.2 Contents The MFC-1 package is used for the transport of fresh fuel assemblies. It is designed to hold up to two Type 14x14 (10 feet [ft]), Type 14x14 (12 ft), Type 15x15 (12 ft), or Type 17x17 (12 ft)

(also including the case where non-nuclear fuel core internals are built in for each type) PWR fresh fuel assemblies. The contents are limited to a Type AF quantity of radioactive (fissile) material.

1.3 Criticality Safety Index The criticality safety index (CSI) for the MFC-1 package containing PWR fuel is 0.

2.0 STRUCTURAL EVALUATION The objective of the structural evaluation is to verify that the structural performance of the Model No. MFC-1 package meets the requirements of the IAEA SSR-6, 2018 Edition.

The NRC staff previously reviewed the structural performance of the Model No. MFC-1 package and recommended revalidation of the Japanese Competent Authority Certificate J/105/AF-96, Revision 2 (ML120030069) that conformed to the 2005 Edition of IAEA Safety Requirements, TS-R-1, Regulations for the Safe Transport of Radioactive Material.

In accordance with the Memorandum of Understanding between the two agencies, DOT requested the NRC to review the Model MFC-1 package and provide a recommendation for upgrading the regulatory citation to the IAEA SSR-6, 2018 Edition, for import and export shipments.

2.1 Description of the Structural Design and the Amendment The Model No. MFC-1 package is a Type A fissile material transportation package and has the following nominal outside dimensions: approximately 5400 millimeters (mm) in length, 1150 mm outside diameter and 1275 mm in height. The total mass of the package is 4320 kilograms (kg).

The MFC-1 package consists of an external cylinder (comprised of an upper cover and a lower container) and a cradle assembly (comprised of a cross frame, shock mount frame, and auxiliary devices). A maximum of two fuel assemblies can be contained in a cross frame and transported in a horizontal position. Brackets are attached on top of the upper cover at four

3 positions for lifting the package during transportation. Wood (acting as both heat insulator and shock absorber) fills the space between the external cylinder and internal cylinder. Borated stainless steels plates act as a neutron absorber and are attached to the entire surface of the cross frame. Clamping frames fasten the support grids of fuel assemblies and the top nozzle during transport. The shock mount frame is mounted inside the package through the shock mounts. There are no components that act as a containment device in the MFC-1 packaging.

The containment boundary consists of the fuel cladding and the end plugs of the fuel rod.

In reviewing the application, the NRC staff noted that there is no physical change to the previously approved package or the package contents. However, the applicant has updated the SAR for the package, primarily adding aging evaluations in the SAR section II-F to comply with paragraph 613A requirement of SSR-6, 2018 edition. In addition, the applicant has made numerous administrative changes to reorganize different parts of the SAR, added supplemental information in the SAR sections II-A.10.9 through II-A.10.12, and made editorial changes to resolve comments from the Competent Authority of Japan.

The NRC staff reviewed the changes in IAEA SSR-6, 2018 Edition, from its 2005 version, Regulations for Safe Transport of Radioactive Material, TS-R-1, Rev. 2005, and finds that paragraph 613A requirement for considering aging mechanisms in the design of packages intended to be used for multiple shipments may have implications to the structural integrity of the affected package components. Other changed requirements in 2018 edition are not applicable, or do not affect the structural design of this package and are not considered for the structural evaluation. Therefore, the scope of structural review remains focused on the review of the aging considerations in the package design, especially for the fatigue mechanism, other updated information included in the applicants SAR.

2.2 Structural Evaluation of the Amendment 2.2.1 Aging Mechanism - Fatigue Paragraph 613A of IAEA SSR-6 requires that aging mechanisms be considered in the design of the package. IAEA Specific Safety Guide No. 26 (SSG-26) provides guidance on how to comply with paragraph 613A in IAEA SSR-6. The Model MFC-1 is a Type A 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 applicant provided a detailed aging evaluation in the SAR section II-F and supplemental information dated October 23, 2024. The NRC staff reviewed and evaluated with respect to the structural integrity of the package components affected by fatigue. Refer to section 7.0, Materials Evaluation, of this SER for detailed evaluations for other aging mechanisms.

The applicant stated that the expected working life for the Model MFC-1 packages is approximately 60 years, the transportation frequency is about six times a year. Based on this frequency, the estimated total number of transport cycles is 360 times during the lifetime.

Each transportation period usually ranges from several days to a month. The applicant has identified reusable package components that are important to safety as follows: external cylinder (upper cover, lower container and tie-bolts), cradle assembly (shock mount frame, cross frame and shock mount), cross frame skin, and shock absorber (i.e., impact limiter). The cross frame skin is borated stainless steel material, it is lightly stressed and not used as

4 structural members during service conditions. There is no repetitive stress caused by handling or other types of service conditions on wood which is used as impact limiter material. Therefore, the fatigue effects in the cross frame skin and the impact limiter material are not evaluated. The external cylinder and cradle assembly components are made of carbon and low-alloy steel. The applicant has evaluated the effects of repetitive stress on the external cylinder, the brackets and the tie-bolts due to the lifting operation and other applicable types of stress cycles.

Handling (Lifting) Fatigue Cycles:

The external cylinder, brackets and tie-bolts are subject to repetitive loads due to handling of the package. In section II-A.4.4 of the SAR, the applicant described and provided a diagram of the lifting device for the package. The external cylinder and brackets are carbon steel, and the tie-bolts are alloy steel. In section II-A.10.9 of the SAR, the applicant has considered 6,000 lifting cycles (15 lifts per transport, 360 transports over the 60-year lifetime) for the fatigue evaluation. The cross-pin for the tie-bolts needs to be replaced within 13 years from the start of service as specified in the applicants internal maintenance manual; thus, the cross-pin is evaluated for a 13-year service life.

The applicant provided that the number of stress cycles considered for the lifting evaluation are lower than the respective allowable number of stress cycles based on the calculated maximum repetitive stresses and the established design fatigue curves in the attachment figures 4.2.1 and 4.2.4 of the request for supplemental information response for the carbon steel and high-tension bolt material. Further, the applicant demonstrated that the allowable number of cycles for the cross-pin associated with the cyclic peak stress intensity is 2,800, which is larger than the number of lifting operations for 13 years (1,170 times).

Based on the review of the applicants evaluation and the relevant SAR sections, the NRC staff finds that the 6,000 lifting cycles considered in the evaluation also includes the evaluation of the tie-bolts due to the tightening torque, which is conservative since the tie-bolts are only installed and removed twice for each transport cycle. Even with the consideration of once-a-year periodic inspection, the total number of the bolt tightening cycles amounts to only 780 cycles (2 x 360

+60) over the lifetime which is less than 6,000 lifting cycles. Furthermore, the NRC staff finds that the referenced design fatigue curves used in the applicants evaluation are appropriate for the material the components are made of, and they are almost identical to figures I-9.1M and I-9.4M, Design Fatigue Curves for Carbon Steels and High Strength Steel Bolts per the 2010 or later American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code (BP&VC), section III, Division 1, Mandatory appendix I. The NRC staff finds the applicants fatigue assessment for the lifting cycles for the bracket, external cylinder and tightening bolts including cross-pin to be acceptable because the estimated number of stress cycles are shown to be less than the allowable number of stress cycles.

The NRC staff notes that the applicant did not evaluate the cradle assembly for any types of fatigue stress cycles. The cradle assembly is supported on both sides from the lower container shell by anywhere from 16 to 20 shock mount supports depending upon type of fuel assemblies.

The cradle assembly components are made of carbon and low-alloy steel material, except for the shock mounts which are polybutadiene rubber material. The NRC staff notes that the cradle assembly is subject to lifting fatigue cycles. However, the estimated lifting cycles are limited to 360 cycles, which is the same as the total number of transport cycles, because the cradle assembly gets loaded with fuel assemblies only once during a transport cycle, and a routine lifting operation of the package does not cause any significant changes to the cradle assembly loading pattern. Also, as stated in the SAR section III-B, Maintenance Conditions, the cradle

5 assembly and the shock mounts are checked by periodic inspection at least once a year for any deterioration or damage, such as cracks and deformations, and the shock mounts are replaced within 13 years regardless of abnormalities found during the inspection. Thus, the NRC staff finds that aging due to fatigue effects is insignificant in the cradle assembly due to the low number of lifting cycles and supplemented with periodic inspection and replacement of components as needed by the maintenance procedure. Thus, the NRC staff concludes that lifting fatigue cycles are not a concern for the structural integrity of the cradle assembly for the estimated lifetime.

Transport Vibration Fatigue Cycles:

The applicant evaluated vibration fatigue cycles in the external cylinder and tie-down bolts and showed that the cyclic stress amplitudes in the external cylinder and tie-bolts are below the fatigue endurance limits based on the 1.0 grams (g) acceleration loads. The applicant selected 1.0 g acceleration loads based on the IAEA SSG-26, table IV.1, Acceleration Values for Strength Analysis, since the mode of transport is by land or the sea.

The applicant provided a depiction of a typical tie-down arrangement in the SAR figure II-A.14 where the package is blocked in the front and rear direction of travel and held down to the truck bed using wire ropes to restrain it in the lateral and vertical directions. Based on this arrangement, the applicant evaluated the external cylinder and tie-bolts due to vibrations during transportation to demonstrate that the resulting cyclic stress amplitude in the cylinder and the tie-bolts remain lower than the fatigue endurance limit of the cylinder and the tie-bolts material.

The NRC staff consider this acceptable based on the cyclic stress amplitude in the most critical element (i.e., tie-bolts), which shows sufficient margin in the material endurance limit of approximately 93 megapascals (MPa) to accommodate up to 2.0 g accelerations from the attachment figure 4-2.4 of the October 23, 2024, supplement. As for the cradle assembly, the NRC staff notes that anywhere from 16 to 20 shock mounts are installed on the cradle assembly supports onto the container shell to reduce the effect of any acceleration or vibration on the fuel assemblies under the routine conditions of transport. Based on these considerations, the NRC staff concludes that the package will withstand vibration normally incident to the transport without any adverse consequences to the structural integrity of the critical package components.

Thermal Stress Fatigue Cycles:

The applicant states that the contents of the MFC-1 package do not generate heat; therefore, the temperature inside is uniform. In addition, since the same material (carbon steel) is used for both the outer cylinder and the dome plates, thermal stress due to different thermal expansion is not a concern. Furthermore, the plate thickness of the external cylinder and the dome plates are only 4.5 mm to 9 mm (i.e., thin sections), so the thermal stress due to the temperature fluctuation of the external environment can be ignored in the container. As for the tie-bolt and cross-pin, the applicant evaluated thermal cyclic stresses due to the variation in the environment temperature and due to the slight difference between the coefficient of thermal expansion between the bolt, the cross-pin and the flange material to demonstrate that the thermal repetitive stresses are relatively small and thermal fatigue is not a concern.

The NRC staff reviewed the applicants assessment and found that the applicants estimated 10,800 thermal cycles for the lifetime is based on 360 transport cycles and 30 days duration for each transport cycle. The calculated cyclic stress in the tie-bolt is about 29 MPa and based on the established fatigue design curves for the tie-bolt material, the allowable number of thermal cycles are 1.0 x 106, which is greater than the estimated thermal cycles for the lifetime. Based

6 on a comparison of the tie-bolts results, the NRC staff notes that the thermal stresses in the cylinder and the dome plates are less critical due to the environmental temperature variations and are negligible. Therefore, the NRC staff concludes that the thermal stress fatigue cycle is not a concern for the structural integrity of the tie-bolts and the external cylinder of the package for the expected service life.

Pressurization Fatigue Cycle:

The MFC-1 package external cylinder is designed for an internal design pressure of 50 kilopascals (kPa) from the SAR section II-A.5.1.1. The stress cycle due to pressure variations is evaluated in the supplement dated October 23, 2024, assuming the temperature variation of the external environment from nighttime (20 degree Celsius [°C]) to daytime (73°C). In this condition, the pressure variation is about 19 kPa, and the resultant repetitive stress is about 6 MPa at the outer cylinder, about 84 MPa at the dome plates and 243 MPa in the tie-bolts cross-pin. The allowable number of cycles corresponding to the maximum cyclic stresses in the cylinder and tie-bolt cross-pin are sufficiently large compared with the expected pressure cycles of about 10,800 for the cylinder and 2,340 for the cross-pin lifetime, respectively, so that fatigue failure does not occur.

The NRC staff reviewed the applicants evaluation and the relevant SAR sections and ensured that the estimated number of pressure cycles considered in the evaluation and calculated repetitive stresses are appropriate considering the day and night temperature differences. The NRC staff also finds that the allowable numbers of stress cycles obtained from the fatigue design curves for the external cylinder and dome plates material are appropriate. The NRC staff finds that the applicants fatigue assessment for the pressure variation inside the external cylinder and the tie-bolts to be acceptable since the estimated number of stress cycles are less than the allowable number of stress cycles based on the established fatigue design curves. The NRC staff concludes that the fatigue cycle due to pressure variation is not a concern for the structural integrity of the package for the estimated lifetime.

Combined Effects of Fatigue Cycles:

The applicant evaluated the combined effects of cycle types in the package components using a traditional method of superposition and then calculating a cumulative usage factor for the applicable component. The applicant shows that the cumulative usage factor for the tie-bolts fatigue strength combining lifting and pressurization cycles is less than 1.0, so the fatigue failure of the tie-bolts will not occur. The applicant also shows the cumulative usage factor for the tie-bolt cross-pin fatigue strength combining lifting, thermal and pressurization cycles is less than 1.0, so the fatigue failure of the cross-pin will not occur.

The NRC staff reviewed the applicants evaluations for combining the effects of cycle types and finds that they are properly combined using an appropriate method similar to that found in the ASME BP&VC section III Code. The NRC staff notes that the usage from the thermal and vibration cycle types are ignored while calculating the cumulative usage factor for the tie-bolts combining the lifting and pressurization cycle effects, and finds it to be acceptable since the usage factors from the thermal and vibration cycle types are negligible, and the calculated cumulative usage factor is less than 1.0 with a design margin. By comparison, the NRC staff finds the cumulative fatigue usage factor for the external cylinder and other components to be less critical and acceptable considering relatively higher allowable numbers for a fatigue cycle type, or low repetitive stresses in components, or that some of the cycle types for these components are not applicable.

7 Based on the above assessments, supplemented with the periodic inspection and maintenance program, the NRC staff does not expect any adverse impact on the structural adequacy of the reusable important to safety package components due to the combined aging effects from various applicable fatigue cycle types.

2.2.2 Other Changes affecting Structural Evaluation In this amendment, the applicant provided an evaluation of the external cylinder due to a maximum external pressure under the cold ambient condition in the SAR section II-A.10.10 and added supplemental information to update the safety margin of the fuel rod cladding design in the SAR section II-A.10.12.

The evaluation of external pressure in the SAR section II-A.10.10 demonstrates that the integrity of the package is not impaired when the ambient environmental temperature varies from 38°C to -20°C. The NRC staff reviewed the evaluation and found that there is margin available in the design when the calculated external pressure load is compared with the external cylinder capacity and, thus, finds it to be acceptable.

In the evaluation of fuel rod cladding under the 9m free drop accident condition of transport, the design is shown to have a safety margin of 0.04 in the SAR section II-A. 9.2.5. The stresses were calculated based on the assumption that the zirconium alloy (cladding material) deforms elastically (i.e. idealized stress-strain relationship exhibiting a straight line curve). The applicant updated the available safety design margin to be 0.2 in the SAR section II-A.10.12 based on the stress-strain diagram of the zirconium alloy obtained from the test data as shown in the SAR figure II-A.74. According to this diagram, the stress generated at the same strain is smaller than that under the assumption of perfect elastic deformation. Based on the review of the updated information, the NRC staff concludes that the updated information has no impact on the applicants previously approved evaluation for the accident conditions of transport.

2.3 Evaluation Findings

Based on the statements and representations in the Model No. MFC-1 package application and as discussed in this SER section, the NRC staff finds that the MFC-1 transportation package meets the structural requirements in IAEA SSR-6, 2018 Edition.

3.0 THERMAL EVALUATION The objective of the thermal evaluation is to verify that the MFC-1 transportation package design satisfies the thermal safety requirements of the IAEA SSR-6 regulations for the safe transport of radioactive material. The NRC staff reviewed the thermal material properties, the descriptions of the thermal modeling, the assumptions used in the thermal analyses, and the calculations provided by the thermal models for normal and accident conditions of transport under the revalidation request for MFC-1 fissile material package with 14x14, 15x15, and 17x17 fuel rod assembly configurations. The purpose of the MFC-1 package is to transport fresh UO2 and Gadolinia-UO2 fuel assemblies made up of enriched natural uranium PWR fuel assemblies.

The MFC-1 is categorized as a Type A package containing fissile material and is required to meet or exceed the IAEA requirements for Normal Conditions of Transport (NCT), both hot and cold, and Hypothetical Accident Conditions (HAC), as described in the thermal requirements of IAEA SSR-6. The design of the package and its contents are described in section I of the SAR

8 for the Model MFC-1 package. The thermal analysis of the package is reviewed in section II-B of the SAR, and details relevant procedures, properties and results.

3.1 Packaging The thermal test, as defined by the IAEA SSR-6 regulations, follows the mechanical tests to account for the most damaging conditions for the MFC-1 package. Subsequent to the HAC mechanical tests, the fuel cladding had not been ruptured. This was confirmed by post-HAC helium leak tests. The applicant determined that the package as modeled showed no deformation in a horizontal position under the NCT while the package is modeled accounting for the deformation under transport during HAC. For the thermal analyses, the applicant employed conservative assumptions regarding the behavior of the fuel rod cladding as a function of temperature. The NRC staff evaluated the following risks in the thermal review:

(1) The temperature rise of the package due to solar insolation.

(2) The temperature rise of the package due to fire testing after significant deformation generated from the drop tests.

The applicant performed three-dimensional analyses of the MFC-1 package with the Three-Dimensional Non-Steady Thermal Analysis Code TRUMP to verify the thermal design of MFC-1 under NCT and HAC in compliance with the regulatory safety requirements of IAEA SSR-6, Revision 1, 2018 Edition. TRUMP is a heat transfer calculation program based on the node method. The analysis model considered model geometry, analysis conditions, and heat transfer methodology of the package.

3.2 Modeling Setup for Confirmatory Analysis The NRC staff reviewed the descriptions and the model calculations of MFC-1, and the thermal properties used in the analyses and evaluated the thermal performance under both NCT and HAC. The parameters and physical phenomena used in the applicants evaluation of MFC-1 under NCT and HAC are summarized in the following subsections.

3.2.1 NCT The thermal properties of the materials are provided in section II-B of the SAR, Thermal Analysis, including those of homogenized fuel, Zircaloy fuel rod cladding tubes, UO2 pellets, balsa wood, carbon steel, and air. The physical properties of the materials (thermal conductivity, specific heat, viscosity, density) vary depending on temperature.

(1) The ambient air is still with a temperature of 38°C (100 degrees Fahrenheit [°F]).

(2) The maximum temperature of the package is 73°C (163°F), and its minimum is -20°C (-

4°F).

(3) The internal pressure rise of the packaging at the maximum temperature is 0.019 MPa.

(4) Leak-tightness can be maintained under NCT.

(5) The decay heat of the content can be negligible.

9 (6) Heat transfer between the package surface and the surrounding environment by natural convection and radiation are taken into consideration.

(7) No deformation of the packaging is observed under NCT.

(8) No thermal stress is produced since there is nothing to restrict the thermal expansion.

3.2.2 HAC The package was exposed to a fire of 800°C (1472°F) or equivalent heat application method for a period of 30 minutes with a calorific value of 0 W, a flame emissivity of at least 0.9, and a surface absorptivity coefficient of 0.8.

The initial temperature of the thermal test is 73°C (163°F) for the whole region of the package, based on the temperature calculation results for NCT. The ambient temperature before the thermal test is 38°C (100°F).

The HAC mechanical tests do not breach the fuel rod cladding which forms the containment boundary. This is confirmed by the post-HAC-test helium leak tests.

The ambient emissivity surrounding the package before, during, and after the HAC thermal test is modeled to be 1.0, 0.9, and 1.0 respectively. The corresponding emissivity at the external surface of the package is modeled to be 0.84, 0.8, and 0.6 respectively.

Heat transfer between the package surface and the surrounding environment by both forced convection and radiation are taken into consideration.

The analysis was performed taking into consideration the deformation of the package produced due to HAC mechanical tests.

The maximum temperature of the fuel rods is 442°C (828°F), which is lower than their allowable temperature limit of 860°C (1580°F). The fuel rods experienced a maximum pressure of 7.82 MPaG in the 17x17 fuel rod bundle, which is below the allowable HAC working pressure of 35.9 MPaG and a maximum stress of 57.4 MPa which is below the stress criterion of 282 MPa.

3.3 Conclusion related to the Thermal Evaluation of Package Material The applicant stated that the maximum temperatures of the MFC-1 package under NCT and HAC are 73°C (163°F) and 442°C (828°F), respectively. The NRC staff reviewed the package model descriptions, input parameters and boundary conditions of the thermal analysis, thermal properties of the packaging material, and verified the temperature distributions under NCT and HAC. The NRC staff concludes that the thermal evaluation is conservative and acceptable for the MFC-1 package based on the conditions considered in the thermal analyses including:

(1) The package is subject to a hot ambient temperature of 38°C (100°F) for NCT testing.

(2) The initial temperature of the package is set as 73°C (163°F) for HAC fire test.

10 (3) The deformation of the package is considered to increase heat conduction into the package for HAC fire test.

(4) The package is exposed to an 800°C (1472°F) fire or equivalent for a period of 30 minutes with a flame emissivity of at least 0.9 and a surface absorptivity coefficient of 0.8 for the duration of the 30-minute HAC thermal test.

(5) The package is exposed to an ambient temperature of 38°C (100°F), in still air and with solar insolation considered, for the duration of the post-fire cooldown.

3.4 Evaluation Findings

Based on a review of the statements and representations contained in the application, the NRC staff finds that the MFC-1 transportation package has an adequate thermal design that meets the requirements for thermal performance outlined in IAEA SSR-6 for the transportation of fuel assemblies during NCT and ACT.

4.0 CONTAINMENT EVALUATION The MFC-1 package has no component as a containment system. The containment system is essentially the fuel assembly rods which provide containment of the radioactive material being transported. The cladding and the end plugs which are components of every fuel rod form the containment boundary for the package. The fuel rod cladding and the fuel rod end plugs are sealed with welding, and there is no penetration.

The applicant demonstrated the integrity of the containment boundary of the fuel rods based on the results of the structural analysis and the thermal analysis of this package under normal and accident conditions of transport, and the test results of the drop test I, the drop test II and the fire test under the HAC of transport with two prototype packages. In particular:

As discussed in Structural Evaluation of this SER, the leak-tightness can be assured without any damage of the fuel rods, under NCT and HAC of transport.

As mentioned in the Thermal Evaluation of this SER, the internal pressure at the maximum temperature of 73°C on the fuel rods could become 3.73 MPaG. The general membrane stress generated in the fuel rods could be 31.1 N/mm2, which is sufficiently smaller than the design stress strength of Zircaloy-4, 239 N/mm2. Thus, the leak-tightness can be maintained.

As discussed in Structural Evaluation of this SER and as demonstrated in the applicants prototype test (refer to SAR section II-F, Test Report of Prototype Packaging for Model MFC-1 Container), the integrity of the containment boundary can be maintained against the drop impact of 9m under HAC of transport.

As discussed in the Thermal Evaluation of this SER, the leak-tightness can be maintained without any damage of the fuel rods under the environment of 800°C for a period of 30 minutes under HAC of transport.

11 5.0 SHIELDING EVALUATION The objective of the shielding evaluation is to verify that the Model No. MFC-1 package design satisfies the external radiation requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 71.47, External radiation standards for all packages. The NRC staff followed the guidance provided in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, published in August 2020. The MFC-1 package is designed to transport unirradiated PWR fuel assemblies. The package is suitable for both sea and road transport.

5.1 Description of the Shielding Design 5.1.1 Packaging Design Features The MFC-1 package has the following nominal outside dimensions: an overall length of 5,400 mm, an outside diameter of 1,150 mm, and a height of 1,275 mm. The total mass of the package is 4,320 kg (2,804 kg packaging plus 2 fuel assemblies). The MFC-1 package consists of an outer shell (comprised of an upper cover and a lower container) and a cradle assembly (comprised of a shock mount frame, mounts, and auxiliary devices). A maximum of two fuel assemblies would be contained and transported in a horizontal position. The outer shell consists of an upper cover and a lower container. The outer shell is a cylindrical, watertight structure, and its outer surface is so structured that rainfall is difficult to accumulate on the package.

Brackets are attached to the upper cover at four positions for lifting the package during transportation. Wood (acting as both heat insulator and shock absorber) fills the space between the external cylinder and internal cylinder. A neoprene O-ring acts as a seal to prevent water from entering the package internals during NCT. The A2 value of this uranium is unlimited.

Engineering drawings are provided in the SAR. A description of the contents is provided in the SAR and coincides with the contents approved on the Japanese Package Design Certificate No.

J/105/AF-96, Rev. 2. The contents allowed by the Japanese certificate are PWR fuel assemblies fabricated either from UO2 or Gadolinia UO2. The allowable U235 enrichment in the fuel assemblies is five weight percent or less.

The applicants SAR provides a description of the packaging which consists of a main body, lifting trunnions, lid and bottom plug, shock absorber, and seals. The fuel assembly rods provide containment of the radioactive material transported. Therefore, the MFC-1 package does not utilize a containment system.

5.1.2 Summary of Maximum Radiation Levels The applicants calculation for routine and NCT, presented in SAR table 11-D.6, shows the results for gamma, neutron and total maximum dose rate during routine transport conditions and under NCT. The NRC staffs past review of the MFC-1 package found that the applicants shielding model conservatively neglected the separation provided by the personnel barrier and much of the shielding provided by the impact limiters. The dose rates, on the surface of the package, and at a distance of 1 meter (m) from the package, are 0.023 millisievert per hour (mSv/hr.) (2.3 millirem per hour [mrem/hr.]) and 0.006 mSv/hr. (0.6 mrem/hr.), respectively.

Dose rate during the NCT is 0.03 mSv/hr. (3 mrem/hr.).

As the MFC-1 package under conditions normally incident to transportation the radiation level does not exceed 2 mSv/hr. (200 mrem/hr.) at any point on the external surface of the package,

12 and the transport index does not exceed 10, the NRC staff finds that the MFC-1 package meets the requirements in 10 CFR 71.47.

5.2 Source Specification The applicant determined both neutron and gamma source terms with ORIGEN-2 modules within the SCALE code system for UO2 fuel assemblies. The enrichment level of fuel assemblies is less than or equal to five weight percent. Because the proposed fuel assembly type falls in a previously evaluated fuel class and the applicants methodology has not changed from the prior revision, the NRC staff finds its continued use acceptable.

5.2.1 Gamma Source The content of the package is UO2 fuel that includes uranium isotopes and their daughter nuclides. The decay gamma radiation was considered in the evaluation of gamma source. The gamma source was evaluated using ORIGEN-2 code and was based on uranium isotopes and five weight percent enrichment for a period of 10 years. The gamma emission rate evaluated for 18 average energy groups from 0.01 to 9.5 MeV, as shown in table 11-D.1 of the SAR, for nuclides shown in table 11-D.2 of the SAR.

5.2.2 Neutron Source Neutron radiation in fuel originates from spontaneous fission, alpha-n reactions, neutrons produced through subcritical multiplication, and gamma-n reactions, consistent from emission of neutron from uranium dioxides fuel assemblies. The neutron source is negligible, and it is much smaller than that of gamma emission rate.

5.3 Model Specification Section D.3 of the SAR includes most of the model description as it pertains to the applicants shielding analysis. Under routine conditions of transport, only the fuel assemblies and the external cylinder of the package are considered in the shielding model. The internal cylinder of the package and the balsa wood between internal cylinder and external cylinder are not used in the model. The NRC staff reviewed the shielding model, which included a comparison of the details in section D.3.1 of SAR, with the drawings in chapter 1. The applicant used the maximum damage to the outer shell during the horizontal drop and deformation in shielding model under NCT and the minimum distance between the fuel assemblies and the package surface used in the shielding model.

The NRC staff performed a confirmatory sensitivity study to evaluate the applicants modeling.

Taking calculational uncertainty into consideration, the NRC staffs study results concluded with an insignificant effect on external dose rates, which confirmed with the applicants determination.

5.4 Shielding Evaluation The applicant performed its shielding analyses with one dimensional transport calculation code ANISN using the model shown in figure II-D-2 of the SAR for one assembly. The dose equivalent rate was obtained by multiplying the results by number of fuel assemblies and the minimum distance between fuel assembly and surface used in the evaluation. The internal cylinder and balsa wood between internal and external cylinders were ignored in the routine

13 conditions of transport model. The shielding for NCT is the same as routine, while considering maximum displacement of the cradle assembly. There is no significant difference between dose rates for both models.

The NRC staff reviewed the description of the MFC-1 package design features related to shielding and the source terms, and the method and instructions for determining the contents.

The NRC staff also reviewed the shielding analyses and performed confirmatory analyses using SCALE/ORIGEN (Version 6.2.4) for source terms and MCNP 6.2 for shielding analysis. The results of the Monte Carlo N-Particle Transport Code calculation showed insignificant dose rates for package surface itself and at 1 m from the package surface, consistent with application results. Based on the assumptions and approximations used in the analyses as presented in the shielding safety analysis and the results of the analysis presented in the application, the maximum dose rates for NCT determine that the reported values were below the regulatory limits in 10 CFR 71.47.

5.5 Evaluation Findings

Based on review of the statements and representations in the application, the NRC staff finds that the MFC-1 transportation package has been adequately described and evaluated to demonstrate that it satisfies the shielding requirements of SSR-6, 2018 Edition.

6.0 CRITICALITY EVALUATION

The applicant requested revalidation of the Japanese Competent Authority Certificate of Approval No. J/105/AF, Revision 1, dated September 13, 2023, for the Model No. MFC-1 transport package. The MFC-1 package is used for the transport of fresh PWR fuel assemblies.

The allowable fuel assembly types remain unchanged in this proposed revision, consisting of 14x14, 15x15, and 17x17 fuel bundles, and have a maximum enrichment of 5.0 wt.% to comply with the requirements of IAEA SSR-6, Revision 1, 2018 Edition.

6.1 Package Description and Evaluation The applicant submitted a revision to the MFC1 package that modified some assumptions in the criticality safety analysis, primarily the use of a new three-dimensional model to increase the precision of the criticality calculations. In addition, the deformation of the packaging was evaluated for routine and NCT, which was found to be bound by the hypothetical accident condition deformation. The NRC staff determined that the applicant conservatively used the damaged configuration for all aspects of their analysis.

The package and the interstitial spacing between the fuel rods were assumed to be fully flooded with water to maximize neutron moderation, as shown in figures II-E.1 and II-E.2 of the SAR.

Mirror reflection of neutrons was used as the boundary condition to represent an infinite array of packages, and results in a CSI of zero. For the spaces between adjacent packages, vacuum conditions were used to maximize the neutron interaction between packages. Borated stainless steel is used in the packages, at a concentration of 1.0 wt.% boron. The allowable fuel assembly types remain unchanged in this revision, consisting of 14x14, 15x15, and 17x17 fuel bundles, and have a maximum enrichment of 5.0 wt.% 235U.

All original calculations had been performed using the KENO V.a Monte Carlo code, which remains an acceptable code, but results in a user having to make some approximations when modeling for the neutron multiplication factor. To address this, the applicant utilized the updated

14 SCALE 6 Monte Carlo code system with the ENDF/B-VII continuous energy cross-sectional library to perform enhanced modeling of the MFC1 package. This included the addition of polytubes and cardboard as packaging materials, ethylene propylene rubber and stainless steel for the cushions, and carbon steel for the clamping frames and pads for the support grid. In addition, with the updated SCALE package, previous assumptions of material densities for the cladding, plugs, and Boral were updated to include materials that were not available (primarily 30Zn) in the previous version of SCALE used by the applicant in the original calculations.

The updated model resulted in the applicant calculating keffs that are more precise than the original calculations, and account for greater model accuracy. For the 14x14 assemblies, the calculated keff + 3 went up slightly from 0.873 to 0.878. For the 15x15 assemblies, the calculated keff + 3 went down slightly from 0.936 to 0.929. For the 17x17 assemblies, the calculated keff + 3 went down slightly from 0.934 to 0.925. In all cases, the calculated keffs remain well below the upper subcriticality limit (USL), which is expected, since although the code was updated, the original calculations were and remain valid.

The NRC staff performed independent confirmatory calculations of the MFC1 package using the SCALE 6.1 package with ENDF/B-VII cross-sections using the 17x17 12-ft fuel assemblies for reference since the delta of the calculated keff was greatest. These calculations used similar assumptions as those utilized by the applicant regarding arrays of packages under HAC, including fully flooded conditions and mirror reflection. The staffs confirmatory calculations agreed well with the applicants calculations, and the maximum keff was well below the USL.

6.2 Evaluation Findings

Based on the review of the competent authoritys certificate, the statements and representations contained in the application, and its confirmatory calculations, the NRC staff finds that the MFC1 package continues to meet the standards of IAEA SSR6, Rev. 1, 2018 Edition.

7.0 MATERIALS EVALUATION The NRC staffs materials evaluation determines whether the applicant adequately described and evaluated the materials used in the MFC-1 package to ensure that the package meets the requirements of IAEA SSR-6. For the review of the 2024 version of the MFC-1 package application against the requirements of the 2018 Edition of IAEA SSR-6, the NRC staffs materials review focused on the updates to the package application to evaluate aging of package components and materials to satisfy the new regulatory requirements in the 2018 Edition of IAEA SSR-6.

7.1 Evaluation of Changes to Comply with New IAEA SSR-6 Requirements Regarding Aging The NRC staffs technical review, documented below, addresses the applicants identification and evaluation of package component materials, loading conditions, service environments, material aging mechanisms, and description of package inspection and maintenance activities for managing effects of credible aging mechanisms. To assist in performing its review, the staff applied technical knowledge and insights, as appropriate, from the NRC guidance in NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, July 2019 (ML19214A111), for managing component aging in spent fuel storage systems.

15 7.1.1 Evaluation of Material Aging Mechanisms to Comply with IAEA SSR-6, 2018 Edition To address the new IAEA SSR-6 paragraph 613A requirement and associated SSG-26 guidance related to component aging, the 2024 version of the MFC-1 package application includes a new section (section II-F) on the evaluation of material aging mechanisms for the MFC-1 package components. This section specifies that the planned period of use of the packaging is 60 years.

Section II-F of the application evaluates the material aging mechanisms associated with the transportation and handling of the loaded package and storage of empty packaging components when not in use. The applicants evaluation of material aging mechanisms in section II F cites the relevant provisions of the package handling procedures and maintenance criteria in section III to address detection of aging effects and repairs and replacements for the MFC-1 package components.

7.1.2 Package Components and Materials To identify and evaluate the potential aging mechanisms, section II-F of the application includes a list of package components and materials that are included in the scope of the aging evaluation. The scope of components subject to aging evaluation includes reusable components of the packaging that perform a safety function. The application identifies the following materials for package components that receive aging evaluation:

The NRC staff confirmed that the scope of the applicants aging evaluation includes all long-lived reusable packaging components that perform a safety function. The materials for these components include borated stainless steel, carbon steel, and wood. The application states that the O-rings, which are not relied upon for containment, are appropriately replaced in accordance with package handling procedures and maintenance criteria. Based on the replacement criteria for O-rings, the application identified that it is not necessary to consider their aging. The staff confirmed that an evaluation of potential aging mechanisms for the O-rings is not needed since they are replaced in accordance with package handling procedures and maintenance criteria.

Therefore, staff found that the applicants identification of package components and materials that are subject to aging evaluation is acceptable since it ensures that materials for long-lived Packaging Components Material Cross Frame (Skin)

Borated Stainless Steel Upper Cover (External Cylinder, Dome Plate, Internal Cylinder, Flange, Bracket, Tie Bolt)

Carbon Steel Lower Container (External Cylinder, Dome Plate, Internal Cylinder, Flange, Shock Mount Attaching Plate, Rib)

Shock Mount Frame (Side Rail, Cross Tube)

Cross Frame (Top Frame, Bottom Frame, Cross Tube, Unibar Channel)

Cross Frame (Clamping Frame, Support Grid Pad, Support Grid Pressing Bolt, Jack Screw, Fixing Frame, Bottom Support, Axle)

Shock Mount Rubber (Polybutadiene Rubber)

Shock Absorber Wood (Balsa Wood)

16 reusable package components are appropriately evaluated for potential aging mechanisms over the service life of the package.

7.1.3 Evaluation of Package Aging Mechanisms The application states that the planned period of use of the MFC-1 package is 60 years with a total number of package transports of 360 during the 60-year period. The application identifies the environmental and loading conditions that are considered in the evaluation of potential aging mechanisms during the 60-year period. These conditions include the highest analyzed temperature (heat) for NCT, radiation emitted by the radioactive contents, chemical reactions in materials that may lead to corrosion of components, and fatigue of package structural components caused by cyclical stress in components due to mechanical lifting cycles, package enclosure pressurization cycles, thermal cycles, and vibration during transport. The applicant evaluated each of these conditions to determine aging mechanisms that could potentially lead to package component degradation during the 60-year period of use of the package 7.1.4 Evaluation of Potential Aging Associated with Heat, Radiation, and Chemical Reactions The application states that, in the evaluation of aging due to heat, radiation, and chemical reactions, 60 years of continuous use of a loaded package is considered as a conservative assumption that bounds the actual planned use of the package. Additionally, the applicant identified a 13-year service life for the rubber shock mount in the evaluation of aging due to chemical reactions. The staff reviewed these assumptions and determined that that they are conservative since package components are only exposed to radiation from the fuel contents during limited intervals for transport and handling of a loaded package. Empty packagings placed in long-term storage indoors would not be exposed to radiation from the fuel contents, and storage temperatures for empty packagings would be limited to values that are less than the highest analyzed temperature for NCT. With respect to chemical reactions, the staff verified that direct exposure of package components to outdoor ambient conditions is limited per the package handling and maintenance criteria; specifically, components are not continuously exposed to outdoor air and water environments during handling and transport of loaded packages, and empty packagings are generally required, per the package handling and maintenance criteria, to be stored indoors or stored outdoors under a waterproof cover when not in use. Therefore, the actual potential for chemical reactions with air, water, moisture, and chemical compounds present in the outdoor ambient environment is bounded by the applicants 60-year continuous use assumption. Based on these considerations, the staff finds that the applicants assumption of 60 years of continuous use of a loaded package is acceptable for the evaluation of aging mechanisms associated with heat, radiation, and chemical reactions.

7.1.5 Evaluation of Potential Aging Due to Heat For the evaluation of potential aging due to heat, the applicant determined that the borated stainless steel and carbon steel packaging components covered in the aging evaluation would not be susceptible to adverse changes from exposure to heat since the highest analyzed temperature for NCT would not cause any adverse changes to the material structure (i.e.,

adverse microstructural or dimensional changes) or material properties.

For the evaluation of potential aging due to heat, the applicant determined that the polybutadiene rubber components covered in the aging evaluation would not be susceptible to adverse changes from exposure to heat since the highest analyzed temperature for NCT is well within the service temperature range of the material. Additionally, the applicant stated that the

17 polybutadiene rubber shock mounts will be periodically inspected and replaced if deterioration is observed.

For the evaluation of potential aging due to heat, the applicant stated that although wood can have a reduction in strength when exposed to high temperature environments, the highest analyzed temperature for normal conditions is well below the temperatures at which thermal decomposition of wood occurs. Additionally, the applicant has widespread use of wood in Japanese packages and testing of wood samples from dispositioned transportation packages has shown no reduction in material performance. For these reasons, the applicant has determined that the wood components covered in the aging evaluation would not be susceptible to adverse changes from exposure to heat.

The staff reviewed the applicants evaluation of potential aging due to heat and determined that the highest analyzed temperature for NCT is not a concern with respect to aging degradation for any of the reusable package component materials. The staffs determination is based on confirming that susceptibility to the adverse effects of aging mechanisms caused by steady-state high temperatures in the subject materialssuch as dimensional changes due to thermal creep in structural alloys, unacceptable reduction in the yield and tensile strength of structural alloys, thermal embrittlement of structural alloys, and changes to the structure and properties of organic materialsoccur at significantly higher temperatures for the subject materials than the maximum temperature for NCT. The staff also noted that the licensees aging evaluation appropriately considers the most adverse condition (e.g., highest temperature) for NCT since material aging occurs due to long-term exposure to normal operating and environmental conditions rather than the more extreme short-duration accident conditions (e.g., fire) that have a low probability of occurrence. Therefore, based on these considerations, the staff finds that the applicants evaluation of potential aging due to heat exposure, and its determination that there would be no adverse changes to the subject materials from exposure to heat, is acceptable.

7.1.6 Evaluation of Potential Aging Due to Radiation For the evaluation of potential aging due to radiation, the applicant determined that all the packaging materials covered in the aging evaluation would not be susceptible to adverse changes from exposure to radiation since the cumulative neutron irradiation over the period of use is at least several orders of magnitude less than the lowest neutron irradiation threshold at which adverse changes to the material microstructure and mechanical properties are known to occur. With respect to the borated stainless steel cross frame skin, the applicant stated that the depletion rate of boron-10 is very small, even after the above 60 years of irradiation, due to the low radiation levels associated with new fuels.

The staff reviewed the applicants evaluation of potential aging due to radiation and confirmed that adverse changes to mechanical properties such as neutron embrittlement and loss of fracture toughness are not a concern for any of these materials since the accumulated neutron fluence over 60 years is at least several orders of magnitude lower than the lowest neutron fluence threshold at which adverse changes to the material microstructure and mechanical properties. The staff determined that the low radiation levels associated with fresh fuels would only have a negligible effect on the boron content of borated stainless over the 60-year period of use. Based on these considerations, the staff finds that the applicants evaluation of potential radiation-induced aging, and its determination that there would be no adverse changes to the packaging materials from exposure to radiation, is acceptable.

18 7.1.7 Evaluation of Potential Aging Due to Chemical Reactions For the evaluation of potential aging due to chemical reactions, the applicant stated that the borated stainless steel components have sufficient corrosion resistance due to the formation of a protective passive film in air environments. For the carbon steel components, the applicant stated that corrosion is unlikely due to these components being painted. Additionally, the steel bolts are coated with an anti-seize lubricant that also protects the bolts from corrosion. The applicant also stated that if any corrosion or abnormality is discovered during periodic visual inspections, appropriate corrective action, such as repair or replacement, is taken. For these reasons, the applicant has determined that the metal components covered in the aging evaluation would not be susceptible to adverse chemical reactions.

For the evaluation of potential aging due to chemical reactions, the applicant stated that the polybutadiene rubber components are not susceptible to general corrosion. However, the applicant stated this rubber is susceptible to oxidation in air, resulting in a hardening effect of the shock mounts. The applicant provided an analysis to show this hardening effect would not have an undue effect on the shock mount performance when the shock mount is limited to a 13-year service life. Therefore, the applicant will replace the shock mounts every 13 years even if no other defects are identified during periodic visual inspections. The staff has reviewed the applicants evaluation of aging of the rubber shock mount and has determined that the applicants proposed time-limited aging analysis and 13-year replacement interval is adequate to ensure the shocks mount will perform their intended safety function during the 60-year period of use.

For the evaluation of potential aging due to chemical reactions, the applicant stated that wood is in a closed environment (between the internal and external cylinders of the outer shell) and is therefore not susceptible to aging due to the lack of oxygen. The staff has reviewed the applicants evaluations of aging of wood components due to chemical reactions and determined that the internal and external cylinders surrounding the wood provided adequate protection from moisture and oxygen that could lead to rot or moisture induced degradation such as warping or cracking.

The NRC staff reviewed the applicants corrosion evaluation for the stainless steel components in a sheltered environment and noted that stainless steel passivity may adequately inhibit general corrosion. This is consistent with the guidance in section 3.2.2.1 of NUREG-2214. But, as described in section 3.2.2.2. of NUREG-2214, stainless steel is susceptible to localized corrosion effects, including loss of material due to pitting and crevice corrosion, when exposed to aqueous air environments. Over extended operating periods, in particular, during numerous package transport operations over a 60-year period, these chemical species may gradually degrade the protective passive oxide film on stainless steel surfaces leading to the formation of pits and crevice corrosion. As described in section 3.2.2.5 of NUREG-2214, stainless steel components require high tensile stress (such as weld residual stress) and exposure to aqueous air environments to be susceptible to the formation of cracks due to chloride-induced stress corrosion cracking (SCC). The staff determined that while the sheltered environment around the cross frame skin could contain aqueous air, the necessary high tensile stress is not present for formation of SCC. Per the above discussion, the staff determined that localized corrosion is a credible aging mechanism for stainless steel components in outdoor environments, during the 60 year period of use, and require that adequate visual inspections performed by qualified personnel using qualified techniques are needed in order to detect and evaluate indications of corrosion so that personnel can reliably determine the need for remedial action, such as repair or replacement of components that show unacceptable indications.

19 The NRC staff reviewed the applicants corrosion evaluation for carbon steel components in all environments and confirmed that the use of paint will help protect against corrosion provided that it remains intact. If the paint becomes damaged or deteriorates during routine use and the components are exposed to aqueous air environments, these components are susceptible to general corrosion and pitting/crevice corrosion. This is consistent with the guidance in sections 3.2.1.1 and 3.2.1.2 of NUREG-2214. The staff evaluated the carbon steel used in the MFC-1 transport package and determined that it is not the type of high strength steel (yield strength greater than or equal to 150 kilo pounds per square inch) that would be susceptible to stress corrosion cracking. This is consistent with the guidance in section 3.2.1.5 of NUREG-2214. Per the above discussion, the staff determined that the paint should be inspected for degradation.

Additionally, if the paint does deteriorate, adequate visual inspections should be performed by qualified personnel using qualified techniques are needed in order to detect and evaluate indications of general or pitting/crevice corrosion so that personnel can reliably determine the need for remedial action, such as repair or replacement of components that show unacceptable indications.

The NRC staff reviewed the applicants corrosion evaluation for alloy steel bolts and confirmed that use of a coating will help protect against corrosion provided that it remains intact. If the coating on the bolts becomes damaged or deteriorates during routine use, these components are susceptible to corrosion. The alloy steel bolts in the assembled packaging components are not likely to experience galvanic corrosion when the alloy steel is in direct contact with the carbon steel components of the package due to the similar electrochemical potentials of these two metals. Per the above discussion, the staff determined that corrosion is a credible aging mechanism for the alloy steel bolts, during the 60 year period of use, and require that adequate visual inspections performed by qualified personnel using qualified techniques are needed in order to detect and evaluate indications of corrosion so that personnel can reliably determine the need for remedial action, such as repair or replacement of components that show unacceptable indications. The staffs assessment of the licensees inspection program is provided in SER section 7.2.

7.1.8 Evaluation of Structural Fatigue Due to Stress Cycles in Package Components In section II-F.1 of the SAR, the applicant included a fatigue evaluation for demonstrating that package components will not be susceptible to fatigue failure due to accumulated stress cycles in components. The applicant states that fatigue of the package due to pressurization cycles is not studied since the package does not have a containment boundary. The only components of the package that the applicant found required a study of fatigue strength were the external cylinder, the brackets, and the tie-bolts due to repeated stress caused by lifting and handling.

The applicants fatigue evaluation, as described in section A.10.9 of the SAR, determined the maximum allowable number of cycles for each component. These values are described in table II-A. 43 of the SAR and are greater than the 6000 lifting cycles (15 lifts per transport) for the 60-year period of use of the package. This fatigue study is covered in more detail in the structural evaluation of this safety evaluation report. The staff notes that these components would also be visually inspected prior to each shipment and during periodic maintenance operations and any damage repaired or replaced as necessary.

The staff reviewed the applicants fatigue evaluation and maintenance operations and determined that it adequately demonstrates that package structural components will not be susceptible to fatigue failure during the 60-year term for the package.

20 7.2 Criteria for Managing Effects of Aging Mechanisms on Package ComponentsSection III of the SAR describes package handling procedures that cover loading operations, inspections before shipment, unloading methods, and preparing empty packages.

Section III-A of the SAR provides pre-shipment inspections, including visual inspections to verify:

The package and contents are free from harmful damage, cracks, or deformations.

The lifting brackets and attachments are free from harmful cracks or deformations.

The boron stainless steel sheets are free from harmful cracks or deformations.

After closing of the package, the does rates on the package surfaces and at 1m away are measured to ensure they are below acceptable limits.

Section III-B of the SAR provides maintenance criteria, including criteria for periodic visual inspection of the packaging components, to ensure the packaging is in good condition over the 60-year period of use. If needed, components will be repaired to maintain proper package integrity. The maintenance criteria specifies that maintenance operations are performed every 10 transports or every year, whichever occurs first.

The maintenance criteria are described in table III-B.1 and include the following:

Visual inspections involving:

The cradle assembly and top cover for harmful damage, cracks, deformation, etc.

O-rings for harmful abrasion, cracks etc. O-rings will be replaced when any abnormal deterioration is confirmed by visual inspection.

Shock mounts for harmful damage, scratches, or elongation. The shock mounts shall be replaced when any abnormal deterioration is confirmed by visual inspection, or every 13 years even if no abnormalities are identified.

Borated stainless steel sheets for harmful cracks, deformation, etc.

Air-tightness test:

Keep the internal pressure of a packaging at air pressure of 40 kPa for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, and inspect the pressure drop.

After keeping it for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the internal pressure of the packaging shall be 30 kPa or more.

Operation check:

Air valves test to ensure proper opening and closing.

Relief valves lifted and tested to ensure proper lift setpoint.

The staff identified that the package handling and maintenance criteria described in section III of the SAR do not include specific provisions for inspection of carbon steel components to detect and evaluate indications of general and localized corrosion to ensure that components with unacceptable general or localized corrosion are repaired or replaced.

The staff identified that the package handling and maintenance criteria described in section III of the SAR do not include any specific provisions for inspection of alloy steel bolts to detect and evaluate indications of general corrosion to ensure that bolts with unacceptable general corrosion are replaced.

21 The staff identified that the package handling and maintenance criteria described in section III of the SAR does not include any specific provisions for inspection of stainless steel components to detect and evaluate indications of localized corrosion to ensure that components with unacceptable localized corrosion are repaired or replaced.

To address the need for adequate inspection, flaw evaluation, mitigative measures, and corrective actions for managing corrosion of metallic components and bolts, the NRC staff issued a RAI, dated May 14, 2025, to include the codes, standards, and/or other methods implemented to ensure that package maintenance activities are adequate to manage the effects of aging in metallic package components that would see long-term use, such that package components are capable of performing their requisite safety functions throughout the period of use.

In response to this RAI, the applicant stated that it has added the following maintenance criteria to its internal periodic inspection procedures:

Visual inspections of:

Painted surfaces of carbon steel components (the upper cover, lower container, cradle assembly and clamping frame) for damage such as peeling. Damage to paint that could lead to corrosion will be repaired by partial or full repainting. Minor corrosion will be removed and the affected area repainted. Components will be replaced if corrosion is found that could affect component performance.

Borated stainless steel sheets (the cross frame skin) for corrosion (local or crevice) that could affect component performance. Corrosion will be treated and repaired as necessary.

Alloy steel tie-bolts for damage of the plating and corrosion. These bolts will be replaced if plating damage, that could lead to corrosion, is found.

The NRC staff evaluated the maintenance criteria provided by the applicant in section III of the SAR and the applicants RAI response, and found the inspections, flaw evaluations, mitigative measures, as well as corrective actions to be acceptable in managing the aging mechanisms identified in section 7.1.1 above.

Based on the above, the NRC staff finds the Aging Management Program, with the additional conditions noted above, is adequate for managing the aging mechanisms identified per the aging management review.

7.3 Evaluation Findings

Based on a review of the statements and representations in the application, the staff concludes that the applicant adequately described and evaluated the materials used in the MFC-1 package, and that the package meets the requirements of IAEA SSR-6, 2018 Edition.

8.0 QUALITY MANAGEMENT SYSTEM The purpose of the quality assurance (QA) (i.e., management system IAEA SSR-6, 2018 Edition) review is to verify that the package design meets the requirements of the IAEA SSR-6,

22 2018 Edition. The staff reviewed the description of the QA program for the Model MFC-1 package against the standards in the IAEA SSR-6, 2018 Edition.

8.1 Evaluation of the Quality Assurance Program The applicant developed and described a QA program for activities associated with transportation packaging for nuclear fuel materials. These activities include design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use. The applicant described the QA organizations independence from other branches in the organization, which includes those responsible for product cost and schedule. The applicants description of the QA program meets the applicable requirements of IAEA SSR-6, 2018 Edition and is based on International Organization for ISO Standardization, Standard No. 9001, Quality management systems Requirements, 2015 Edition, and other applicable standards. The NRC staff finds the QA program description acceptable as it allows implementation of the associated QA program for the design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use of the Model MFC-1 transportation package.

Based on the review, the NRC concludes with reasonable assurance that the QA program for the MFC-1 transportation packaging: (1) meets the requirements in IAEA SSR-6, 2018 Edition, and (2) encompasses design controls, materials and services procurement controls, records and document controls, fabrication and maintenance controls, nonconformance and corrective actions controls, an audit program, and operations or programs controls, as appropriate.

8.2 Evaluation Findings

Based on review of the statements and representations in the MFC-1 transportation packaging application and as discussed in this SER section, the staff finds that the MFC-1 package has a quality management system that meets the requirements in IAEA SSR-6, 2018 Edition.

CONCLUSION Based on the statements and representations contained in the documents referenced above, and for the reasons stated in this SER, the NRC staff concludes that Model No. MFC-1 package meets the requirements of the IAEA SSR-6, Revision 1, 2018 Edition. The NRC recommends revalidation of Japanese Certificate of Competent Authority J/105/AF, dated September 13, 2023, for the MFC-1 transport package with the following condition:

The U.S. revalidation term is limited to 5 years.

Issued with letter to Ryan J. Vierling, Chief, Sciences Branch, Office of Hazardous Materials Safety, DOT, dated September 5, 2025.